ML051940267

Response to Commitments from Generic Letter 2003-01, Control Room Habitability

ML051940267
Person / Time
Site:	Palisades
Issue date:	07/07/2005
From:	Harden P Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GL-03-001
Download: ML051940267 (7)

Text

NM C Palisades Nuclear Plant Committed to Nuclear Excll nce Operated by Nuclear Management Company, LLC July 7, 2005 10 CFR 50, Appendix A U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Palisades Nuclear Plant Docket 50-255 License No. DPR-20 Response to Commitments from Generic Letter 2003-01. "Control Room Habitability" By letter dated August 7, 2003, Nuclear Management Company, LLC (NMC) provided a 60-day response to Generic Letter (GL) 2003-01, for the Palisades Nuclear Plant (PNP),

which provided the basis for acceptability, and the schedule for completion of the alternative course of action. By letter dated November 25, 2003, NMC provided specific commitments related to GL 2003-01 and a schedule for completion for the PNP. By letter dated November 23, 2004, NMC provided a revised schedule of commitments for GL 2003-01 for the PNP.

NMC is providing responses to commitments made by the November 23, 2004 letter. describes how the commitments were fulfilled.

Summary of Commitments This letter describes the fulfillment of three commitments made in the November 23, 2004 letter. Commitment #4 to develop Technical Specification changes (and any associated plant modifications) to support requested information remains applicable.

This letter contains two new commitments and no revisions to existing commitments.

NMC will submit a license amendment request for full scope implementation of the alternative source term (AST) methodology on or before July 31, 2006.

27780 Blue Star Memorial Highway. Covert, Michigan 49043-9530 Telephone: 269.764.2000

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Document Control Desk Page 2 NMC will implement the modifications identified and credited in the full scope AST submittal by the end of Refueling Outage 19.

Paul A. Harden Site Vice President, Palisades Nuclear Plant Nuclear Management Company, LLC Enclosure (1)

CC Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC Document Control Desk

ENCLOSURE 1 GENERIC LETTER 2003-01 PALISADES NUCLEAR PLANT RESPONSE TO COMMITMENTS

Background

Generic Letter 2003-01, uControl Room Habitability," requested licensees to verify that the most limiting unfiltered inleakage into the control room envelope (CRE) is within the limits assumed in the safety analysis for radiological control room habitability (CRH),

and that the most limiting unfiltered inleakage value is incorporated into the hazardous chemical analysis. In addition, an assessment of the effect on CRH for internal and external smoke events was requested.

Palisades Nuclear Plant (PNP) has a positive pressure control room and is required by Technical Specifications to maintain greater than or equal to 0.125 inches water gauge during emergency mode operation. The baseline tracer gas testing of the CRE was performed to quantify any unfiltered inleakage through walls, ceilings, floors, doors, and penetrations during a postulated emergency mode operation of the control room heating, ventilation and air conditioning (HVAC) system. The CRE at PNP consists of the mechanical equipment room, the control room, the viewing gallery and the technical support center. The basic test methodology used for the tracer gas tests was the constant injection method of American Society for Testing and Materials (ASTM) consensus standard E741, "Standard Test Method for Determining Air Change in a Single Zone by Means of a Tracer Gas Dilution." This test establishes conditions within the CRE, consistent with those assumed in the accident analyses, for the entire period following a design basis loss-of-coolant accident (LOCA). At PNP, the current design basis inleakage value assumed in the safety analyses is 85 cfm.

NMC is providing responses to three commitments made by letter dated November 23, 2004, as follows:

Commitment 1

1.

Perform the ASTM E741 testing [T] and, [provide] the requested response to GL 2003-01 item 1 (a):

1(a)

That the most limiting unfiltered inleakage into your CRE (and the filtered inleakage applicable) is no more than the value assumed in your design basis radiological analyses for control room habitability.

Describe how and when you performed the analyses, tests, and measurements for this confirmation.

Nuclear Management Company, LLC (NMC) Response

1.

CRE leakage testing was performed April 6, 2005, through April 8, 2005, at PNP.

NUCON International, Incorporated was contracted to perform the test. The test was performed using Palisades Special Test Procedure T-336, uControl Room Envelope Integrated Unfiltered Inleakage Test," Revision 0. The amount of unfiltered air inleakage into the pressurized CRE was determined using NUCON Procedure 12-366, 'Envelope Leakage Testing and Characterization using the Page 1 of 5

Constant Injection Method," Revision 1. The test procedures are based on ASTM E741, to ensure compliance with the requirements of Generic Letter 2003-01. Inleakage for the control room emergency HVAC Trains 'A" and "B" alone, and in dual operation mode, was measured using the constant injection method. The results are as follows:

Unfiltered Uncertainty (2 )

Test Configuration Inleakage (scfm)

(scfm)

Train 'A" Emergency mode with single 13

+/-10 active failure(1)

Train UB" Emergency mode with single 49

+/-9 active failure(1) 49 9

Dual Train Emergency mode with single 124

+/-23 active failure(1

_)_

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+/-23 (1) Most leak resistant normal intake damper, as determined by T-336, failed in the open position.

(2) Per Regulatory Guide (RG) 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Section 1.4, uncertainty is not required to be reported for measured inleakage values less than 100 cfm but is included for information only.

The limiting measured CRE unfiltered inleakage, including bounding uncertainties, is 58 scfm. This value is less than the unfiltered inleakage value of 85 scfm used in the current design basis analyses, as described in the PNP Final Safety Analysis Report (FSAR). However, the analyses of radiological consequences at PNP, are currently non-conforming. This non-conformance is because the methods used to demonstrate the control room radiological design basis do not conform to the latest regulatory guidance.

By letter dated August 11, 2000, PNP made a commitment to update the radiological consequence analyses in accordance with the pending (now issued) regulatory guidance. Numerical Applications, Inc. was contracted to perform radiological analyses for PNP. Re-analysis was performed in 2005, with methodologies conforming to RG 1.183, 'Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," dated July 2000, RG-1.194, 'Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," dated June 2003, and RG-1.196, 'Control Room Habitability at Light-Water Nuclear Power Reactors," dated May 2003. Considering recent NRC interpretations of this guidance, the re-analysis resulted in decreased margin to control room dose limits. Based on preliminary analyses in 2004, the decreased margin is more pronounced when the traditional technical information document (TID)-14844, source term methodology of RG-1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," dated May 2003, is used.

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The documented re-analysis utilized ORIGEN core inventories, RG 1.183 source term, RG 1.194 (ARCON96) atmospheric dispersion factors (x/Q), and appropriately bounding design basis assumptions for all FSAR Chapter 14 analyses that require radiological evaluation. Normal unfiltered inflow is assumed to be 384.2 cfm for the first 1.5 minutes of the design basis accident (DBA) maximum hypothetical accident - loss-of-coolant accident (MHA-LOCA).

At 1.5 minutes, the ventilation system is automatically aligned for emergency (filtered) mode, with only one train operational due to the assumed single active failure of one emergency diesel generator. Note that although inleakage is greater for dual train operation, the limiting assumed inleakage in single train mode is bounding. This is due to the reduced filtered recirculation flow afforded in single train mode. Single train emergency mode intake (filtered) is 1413.6 cfm, and filtered recirculation flow is 1413.6 cfm. Unfiltered inleakage in a range from 0 cfm to 100 cfm is supported by the analyses.

The primary aspects of the conforming methodologies responsible for the decreased margin, relative to the existing design basis analyses are:

• Use of ARCON96 (RG 1.194) for control room x/Q versus use of N/Q derived from site-specific wind tunnel testing

• Use of constant control room breathing rate (RG 1.183, RG 1.195) versus using reduced, time-dependent breathing rates comparable to allowed offsite breathing rates

• Use of engineered safeguards (ESF) room emergency core cooling system (ECCS) leakage iodine airborne fraction of 10% (RG 1.183, RG 1.195) when less than 10% of leaked fluid flashes, versus use of iodine airborne fraction equivalent to leaked fluid flashing fraction

• Consideration of effect of safety injection refueling water tank (SIRWT) pH on iodine re-evolution due to increased iodine re-volatilization versus consideration only of sump fluid pH

• Consideration of all dose significant nuclides with respect to shine dose from SIRWT versus consideration of only nuclides that also contribute to inhalation/immersion dose with respect to shine dose from SIRWT The use of conforming methodologies, in combination with more realistic (less overly conservative) assumptions for source term, generation, removal, and transport of radionuclides, provides reasonable assurance that the design basis embodied in general design criteria (GDC) 1, 3, 4, 5 (5 since Palisades is a single unit site) and 19, can be met without reliance on compensatory actions or measures (e.g., operator ingestion of potassium iodine, operator donning of SCBA.

The decrease in margin, resulting from conformance to the new methodologies, is such that full qualification, with respect to the GDC requirements, cannot be demonstrated without plant modifications to address the dose consequences.

This is due to the SIRWT leakage, SIRWT direct shine, ESF room leakage, and containment leakage components of overall dose. The use of the alternative source term methodology of RG 1.183, and the use of the QAD shielding code, are necessary to allow for the most optimal set of modifications to be pursued.

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Plant modifications and NRC acceptance of alternative source term analyses are necessary in order to demonstrate full conformance with the radiological design bases at PNP. NMC plans to determine the optimal conceptual modification(s) by the end of 2005. Modifications being considered include:

• Relocation of normal intakes and addition of shielding beneath the SIRWT

• Installation of bubble tight dampers on the normal intakes, relocation of the SIRWT vent, and additional shielding beneath SIRWT

• Installation of bubble tight dampers on the normal intakes and relocation of SIRWT tank A submittal of design basis analyses utilizing alternative (RG 1.183) source terms, with credit taken on the intended design, will be made by July 31, 2006.

Upon approval of the analyses and completion of the concurrent detailed design work, implementation of the modifications will be completed by the end of Refueling Outage 19, currently scheduled for fall of 2007.

Commitment 2

2.

Verifying by ASTM E741 testing that the most limiting inleakage has been incorporated into the hazardous chemical assessments (GL 2003-01 item 1 (b) part 1):

1(b)

That the most limiting unfiltered inleakage into your CRE is incorporated into your hazardous chemical assessments.

NMC Response

2.

The current PNP hazardous chemical analyses assume no credit for the control room or the control room ventilation system. The basis credits only atmospheric dispersion between source and receptor located at the control room. Therefore, in this sense, the measured inleakage results are bounded by the existing analyses. However, because the analyses are not in complete conformance with Regulatory Guide 1.78, "Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release," dated December 2001, and because a significant amount of time has passed since the original analyses, additional analyses have been performed.

Numerical Applications, Inc. performed the hazardous chemical assessment and analyses at PNP. The analysis considers a greater number of hazardous chemicals than previous analyses and utilizes the HABIT code to assess CRH.

In the analyses, no credit is taken for control room isolation or purging. The control room air exchange rates measured in the tracer gas test are bounded by those assumed in the hazardous chemical analysis.

The hazardous chemical evaluation was performed in accordance with guidance provided in RG 1.196. Per RG 1.196, the evaluation should review CRE habitability for toxic gases in accordance with RG 1.78. Per RG 1.78, onsite Page 4 of 5

II hazardous chemicals and offsite stationary and mobile hazardous chemicals, within a five-mile radius of PNP, were identified and screened. Detailed analysis of all chemicals that did not screen out was performed.

Based on surveys and other documents, there are a total of approximately 75 potentially hazardous chemicals located, or passing (mobile, along Interstate Highway 1-196), within 5 miles of the PNP control room. Of the approximately 75 chemicals screened, 10 required further evaluation for impact on CRH. Sodium hydroxide, sulfuric acid, and hydrazine are present on-site in quantities large enough to require explicit evaluation. Propane, ammonia, toluene, acetonitrile, iodine, sodium hydroxide, and sulfuric acid are located or transported off-site in quantities large enough to require explicit evaluation.

None of the chemical release scenarios explicitly analyzed exceeded immediately dangerous to life and health (IDLH) limits. All other hazardous chemicals screened out based on the criteria outlined in RG 1.78. Therefore, there are no on-site, or off-site, stationary or mobile hazardous chemical sources that threaten CRH. Tracer gas testing confirms bounding air exchange rates of the PNP CRE are used in the chemical hazards analyses.

Commitment 3

3.

[Perform a] smoke assessment (GL 2003-01 item 1(b) part 2):

1(b)

Also, confirm that the reactor control capability is maintained from either the control room or the alternate shutdown panel in the event of smoke.

NMC Response

3.

Per RG 1.196, the habitability assessment for smoke, and products of combustion, should be performed in accordance with Appendix E of NEI 99-03, "Control Room Habitability Guidance," Rev. 0. The assessment items of the NEI guidance were addressed in an engineering analysis to document the qualitative assessment of smoke events at PNP.

There were no items identified that would impede habitability or prevent operation of the alternate shutdown panels and controls for a smoke event originating in the mechanical equipment room, the control room, the technical support center or the control room viewing gallery. For smoke events external to the CRE, the assessment showed that either the control room, or the alternate shutdown panel, would remain accessible. Smoke from a single credible onsite smoke event will not enter both the CRE and alternate shut down panel area. In the event of minor smoke intrusion, operators may use SCBAs, shutdown the control room HVAC systems, and/or evacuate the control room. Plant operators would be able to achieve and maintain safe shutdown (reactor control capability) from either the control room or the alternate shutdown panels, if needed.

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