ML20077F141

Application for Amends to Licenses DPR-43 & DPR-60,revising TS 3.8 to Allow Containment Airlock Doors to Remain Open During Core Alterations Provided Certain Conditions Met

ML20077F141
Person / Time
Site:	Prairie Island
Issue date:	12/05/1994
From:	Wadley M NORTHERN STATES POWER CO.
To:
Shared Package
ML20077F134	List: ML20077F130 ML20077F141 ML20077F145
References
NUDOCS 9412130276
Download: ML20077F141 (12)

Text

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4 UNITED STATES NUCLEAR REGULATORY COMMISSION 1

NORTHERN STATES POWER COMPANY PRAIRIE ISIAND NUCLEAR CENERATING PIANT DOCKET NO. 50-282 50-306 REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR-42 & DPR-60 LICENSE AMENDMENT REQUEST DATED December 5, 1994 Northern States Power Company, a Minnesota corporation, requests authorization for changes to the Prairie Island Operating License and Appendix A as shown on the attachments labeled Exhibits A, B, and C. Exhibit A describes the proposed changes, reasons for the changes, and the supporting safety evaluation /significant hazards determination. Exhibit B contains current Prairie Island Technical Specification pages marked up to show the proposed changes. Exhibit C contains the revised Technical Specification pages.

This letter contains no restricted or other defense information.

NORTHERN STATES POWER COMPANY By /u b M. D. Wadley Plant Manager Prairie Island Nu ear Generating Plant On this y of Iheforemeanotarypublicinandforsaid County, personally appeared W. D. Wadley, Plant Manager, Prairie Island Nuclear Generating Plant, and being first duly sworn acknowledged that he is authcrized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the statements made in it are true and that it is not interposed for delay.

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l MARCIA A. LeCORE I NOTARY PUBUC.M!NNESOTA l ,

HENNEPIN COUNTY

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LICENSE AMENDMENT REQUEST DATED December 5, 1994 Channes in' Containment Refueline Intearity Reouirements EXHIBIT A Description of the Proposed Changes, The Reasons for Requesting the Changes, and the Supporting Safety Evaluation /Significant Hazards Determination Pursuant to 10 CFR Part 50, Sections 50.59 and 50.90,.the holders of Operating Licenses DPR-42 and DPR-60 hereby propose the following' changes to the Facility Operating Licenses-and Appendix A, Technical Specifications:

Background

Technical Specification 3.8.A, " Core Alterations", currently requires one airlock door closed and one isolation valve on each penetration to be OPERABLE or locked closed durin5 CORE ALTERATIONS. This license amendment request proposes to allow containment airlock doors and penetrations to be open during CORE ALTERATIONS under specified conditions. This request is related in some respects to the previously submitted License Amendment Request dated July 13, 1992, " Refueling and Fuel Handling Specification Changes",

which was subsequently withdrawn. -

Extensive work is conducted in containment throughout a refueling outage -

including the time during CORE ALTERATIONS. In a typical outage day while CORE ALTERATIONS are in progress, there may be 250 personnel entries or exits through the containment airlocks. The airlock doors were not designed for this high level of traffic and damage to the door seals, hingas and interlocks has occurred. In an effort to minimize the damage, trained, dedicated personne1' have been assigned full time during CORE ALTERATIONS to oporate the airlock doors. Nonetheless, the continued high use of these doors continues to require unwarranted door maintenance and may degrade the ability of the doors to perform their function of maintaining containment integrity if a fuel handling accident were to occur. This amendment request proposes to allow the airlock,

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to remain open during CORE ALTERATIONS provided specifie/. conditions are met.

This requested' amendment will allow containment penetrations to remain open under specified conditions. Under existing Technical Specifications significant control room operator resources are expended to control ,

containment penetrations in preparation for.and during refueling. These  :

resources could be more effectively applied' to other outage activities with  !

more safety significance such as decay heat removal related activities. l Furthermore, many of the containment penetrations which provide a direct path- i from containment to the outside atmosphere are small vent, drain and  !

, instrument lines which have small cross-sectional area and would not allow significant releases following a Fuel Handling Accident. Controlling these Page 1

penetrations requires an effort disproportional to their safety significance and' detracts from other more important tasks in the control room. Also, suspending work which could result in small penetrations being open during CORE ALTERATIONS may result in significant cost due to extanding the outage with a negligible benefit in avoided potential releases.

Proposed Changes and Reasons for Channes The proposed changes to Prairie Island Operating License Appendix A, Technical Specifications are described below, and the specific wording changes are shown in Exhibits B and C.

1. Technical Specification 3.8 REFUELING AND FUEL HANDLING. A. Core Alterations. Paracraoh 1.a:

Current Specification 3.8.A.1.a would be replaced as shown in Exhibit B.

Justification: The proposed new Specification 3.8.A.l.a.1, 2 and 3 are consistent with existing Prairie Island Technical Specifications for containment penetrations during CORE ALTERATIONS, except that provisions for allowing airlock doors and penetrations to remain open are included.

Core alteration containment isolation specifications are provided to minimize releases following a fuel handling accident. The proposed amendments will continue to maintain fuel handling accident releases below the current licensing bases and well within the limits of 10CFR100.

Allowing both airlock doors open during CORE ALTERATIONS will facilitate evacuation of containment following a fuel handling accident and help maintain the seals in good working order.

A fuel handling accident does not have an associated mechanism for containment pressurization, consequently releases from an Fuel Handling Accident would be very low. Assuming a single failure results in a supply duct damper remaining open, the containment purge supply fan operating at the time of a fuel handling accident would continue to force air into containment Under these circumstances the containment purge supply fan provides a driving force to push radioactive materials out of containment into the outside atmosphere.

To assure that offsite releases are maintained well within the guidelines of 10CFR100, the proposed License Amendment Request requires the containment (high flow) purge system to be isolated and the inservice (low flow) purge system able to be automatically isolated.

The proposed Technical Specifications impose two additional prerequisites on open containment airlocks. There shall be at least one airlock door OPERABLE which means one door is in working order and Page 2

capable of being closed. Also the airlock door shall be under direct

' procedural control so that it can be closed as soon ac practical following an accident to assure that accident releases are minimized.

With the same containment purge system requirements as the airlocks, containment penetrations are allowed open to free operators to attend to more safety significant outage activities. The existing requirements for control room operators to maintain control over containment penetrations distracts the operators from more important activities. As shown below, the releases with containment penetrations assumed open are well within the limits of 10CFR100 using conservative assumptions, including the assumption the inservice (low flow) purge supply fan continues operating.

Standard Technical Specification wording has been incorporated into Specification 3.8.A.1.a in that "outside" was changed to "outside atmosphere". Releases from containment that are not into the Auxiliary Building Special Ventilation Zone are considered released to the outside atmosphere. Outside atmosphere releases are not filtered and are assumed to migrate to the site boundary unhindered. For the purposes of this amendment request it was assumed that penetrations may be open to the Auxiliary Building Special Ventilation Zone but no credit was taken for filtration by the Auxiliary Building Special Ventilation System. The proposed Technical Specification imposes the same prerequisites prior to opening penetrations to the Auxiliary Building Special Ventilation Zone, that is, the containment (high flow) purge system is isolated and the inservice (low flow) purge system is capable of automatic isolation.

2. Technical Specification 3.8 REFUELING AND FUEL HANDLINC. A. Core Alterations. Paragraph 3:

Current Specification 3.8.A 3 would be replaced as shown in Exhibit B. ;

lustification: Existing Technical Specification 3.8.A.3 requires the l conditions of Specification 3.8.A.l.a to be met if a residual heat i removal pump unexpectedly goes out of service.

This license amendment request proposes to revise Specification 1 3.8.A.l.a to allow air lock doors and containment penetrations open  ;

during CORE ALTERATIONS. During these activities the potential exists for a Fuel Handling Accident for which the consequences are shown below to be well within the limits of 10CFR100. However, during degraded residual heat removal conditions, the potential exists for an accident more serious than a Fuel Handling Accident. Therefore the l provisions of Specification 3.8. A.1.a as proposed in this license amendment request are not appropriate.

This request proposes to revise paragraph 3.8.A 3 to apply the requirements of existing paragraph 3.8.A.l.a, that is, at least one air lock door in each personnel airlock shall be closed and actions Page 3

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'or locked closed in each line which penetrates containment and provides a direct path from containment atmosphere to the Auxiliary Building Special Ventilation Zone or to the outside atmosphere. Under the provisions of the proposed amendment the equipment hatch will already be closed. Through these specifications more stringent containment integrity requirements are applied and provisions are made  ;

to maintain compliance with the Technical Specifications at all times. ]

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3. Technical Specification Bases 3.8 REFUELING AND FUEL HANDLING:

An additional paragraph would be inserted as shown in Exhibit B Justification: A new paragraph would be inserted in the bases to provide background for the changes proposed to Technical Specification 3.8.A.l.a.

As discussed below, these requested amendments for containment airlocks and penetrations to remain open during refueling will maintain accident releases well within 10CFR100 limits, s.

Safety Evaluation The containment serves to contain fission-product radioactivity that may be released during an accident such that offsite radiation exposures are maintained well within the limits of 10CFR100. Technical Specification 3.8.A exists to control radioactive releases from a postulated fuel handling accident in containment.

Ilowever, as a practical matter, Technical Specification 3.8.A.l.a does not prevent all releases from containment to the outside atmosphere due to the necessity to open the containment airlock doors to evacuate the containment workers. Each time the airlocks are opened following a fuel handling accident some radioactivity would be released to the outside atmosphere. Also the workers would be exposed to the Post-fuel handling accident atmosphere in containment while wait 1ng to exit containment. In this circumstance, the Shift Manager could reasonably make the judgement that invoking 10CFR50.54(x) would be the best method to protect the public health and safety and order the airlock open until containment was evacuated. Either way some small amount of radioactive material would be released.

Fuel handling accidents in containment at Prairie Island are addressed in USAR Section 14.5.1 and submittals to the NRC dated March 21, 1977 and January 3, 1979. The USAR considers two accident scenarios, first it considers a sequence involving a fuel handling accident while the low flow containment purge is operating which subsequently isolates due to high radiation sensed in the Shield Building Stack and second it cor iders a worst-case fuel handling accident while the high flow containment purge is operating and fails to isolate.

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The first accident is bounded by the fuel handling accident in the Spent Fuel Poo'l. This event was first evaluated using Prairie Island. specific parameters and assumptions, and a second time using Regulatory Guide 1.25 assumptions. In accordance with the requirements of current Technical Specification 3.8. A.1.a, the containment airlock doors are assumed to be closed and all' releases through the containment low flow purge system would be filtered prior to isolation.

The second and worst-case accident scenario discussed in USAR Section 14.5.1.4 assumes that the high flow containment purge system fails to isolate and all of the radioactive material released from the refueling pool is assumed to be immediately released (puff release) from containment via the high flow  ;

containment purge system to the outside atmosphere. Using conservative assumptions in accordance with Regulatory Guide 1.25, this analysis determined that the exclusion area boundary doses are less than 150 Rem-Thyroid which is within the requirements of 10CFR100.

She NSP submittals to NRC dated March 21, 1977 and January 3,1979 used assumptions similar to the USAR worst-case analyses except that a smaller value for the atmospheric dispersion factor was used. These analyses conservatively determined the site boundary dose to be 82 Rem-Thyroid.

The NRC in its Safety Evaluation Report transmitted to NSP by letter dated February 2, 1982, approved the Prairie Island fuel handling accident based on NRC evaluations using similar assumptions, except that a slightly higher atmospheric dispersion factor was assumed and correspondingly higher doses were calculated to be 102 Rem-Thyroid for a single damaged assembly.

A new analysis for a fuel handling accident was performed in support of this License Amendment Request assuming that the inservice (low flow) purge system is operating when the fuel handling accident occurs and fails to isolate. The ,

offsite dose results of this new analysis are well within the limits of l 10CFR100, as described in Standard Review Plan 15.7 4, Revision 1. This I I

analysis shows that using very conservative assumptions with accident radioactive materials released from containment exponentially the two hour l exclusion area boundary dose is approximately 51 Rem Thyroid. The Standard 1 Review Plan guidelines are 25% of the 10CFR100 liuits, that is, 75 Rem-Thyroid and 6 Rem-Whole Body. Only thyroid doses are given since the whole body doses are far less limiting. The details of this analysis are described in attachment 1, 1 The dose consequence for control room operators was also evaluated as described in Attachment 2 which shows that the two day dose would be 2.7 Rem-Thyroid which is well below 30 Rem which has been determined to be the thyroid equivalent of the 10CFR50 Appendix A, Criterion 19 whole body dose limit of 5 Rem. j l

The potential for containment pressurization as a result of a fuel handling  ;

accident is not likely. In the unlikely event the inservice purge supply fan  !

failed to isolate, operators would manually close supply duct dampers in accordance with alarm response procedures and shut off the fan. Consequently, the amount of radioactive material released through the containment airlocks Page 5

l dur,ing containment evacuation and through open penetrations would be very small.

The containment airlock doors would be under procedural control and one door is required to be OPERABLE at all times during CORE ALTERATIONS. Timely l closure of an airlock door will assure that accident releases are minimized l and actual doses in the event of a fuel handling accident are reduced.

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( The radioactivity released through the airlocks during evacuation following a l fuel handling accident would be offset by reduced worker exposure while

! evacuating containment and increased airlock door reliability. With the high l level of airlock door usage and resulting maintenance, the ability of the

! airlock to seal containment could be compromised. Keeping both airlock doors open during CORE ALTERATIONS could improve the reliability of the airlock doors.

The requested license amendments also propose to allow containment penetrations with direct access to the Auxiliary Building Special Ventilation Zone or to the outside atmosphere to be open during CORE ALTERATIONS. For the analyses in support of these requested amendments, all open penetrations were assumed to be open to the outside atmosphere with no credit taken for the Auxiliary Building secondary containment features.

Radiological consequences due to releases through open penetrations are also bounded by the conservative analyses described in Attachment 1 and 2. The releases would be flow limited at 6000 CFM by the assumed flow of the un-isolated low-flow containment purge supply fan. Following containment evacuation and airlock door closure, releases are assumed to continue to flow through open penetrations.

Actual fuel handing accident releases would be anticipated to be much lower.

As discussed above, if the low-flow containment purge supply fan failed to isolate, the operators would isolate it and remove the driving force for containment releases. Additionally operators would likely initiate the containment cleanup system. As part of post accident emergency actions, the plant staff would also likely identify major penetration openings and take rctions to isolate them if the release consequences outweighed the worker exposure. As stated previously, all of the 6000 CFM release from containment is assumed to go to the outside atmosphere.

If a fuel handling accident were to occur the expected radioactive releases would be only a small fraction of the conservative analyses described in Attachment 1 and 2. The inservice purge system would isolate following closure of the duct isolation valves on high radiation following a fuel handling accident, removing the source for containment pressurization and consequently very little radioactive material will escape from open penetrations.

An analysis of the fuel handling accident inside containment based on expected conditions would show a significantly lower dose consequence. The fuel handling accident was assumed to occur at the earliest time possible following reactor shutdown, 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />; if the fuel handling accident were to occur at 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> the dose would be about 0.7 times that calculated. The radioactive Page 6 f

l iodine in the gap is conservatively estimated; all fuel rods in the dropped assembly.are assumed damaged. More realistic assumptions would decrease the dose proportionately. The fission product inventory is calculated on the basis of the highest power fuel assembly; the dose resulting from an averaSe assembly would be 0.6 times that calculated. It has long been argued that the decontamination factor for iodine in water exceeds 100 as assumed in these analyses; a more realistic factor would decrease the dose proportionately. No credit has been taken for the containment cleanup system which would be manually initiated to filter the containment atmosphere. The worst case meteorological conditions were assumed; the dose during average conditions would be 0.01 times that calculated.

Combining the quantified conservatism identified in the previous paragraph, the expected thyroid dose from the entire fission product release from damaged rods for the case of a Fuel Handling Accident insido containment while operating the low flow purge system is less than 1 Rem-Thyroid without taking any credit for the automatic isolation system.

Thus, the radioactive material expected to be released from the containment airlock doors during containment ovacuation and through open penetrations would be a small fraction of the materials assurued to be released in the supporting analyses and the doses would be a small fraction of 10CFR100.

In conclusion Northern States Power believes there is reasonable assurance that the health and safety of the public will be protected by the proposed Technical Specification amendments.

Determination of Significant Hazards considerations The ptoposed changes to the Operating License have been evaluated to determine whether they constitute a significant har.ards consideration as required by 10 CFR Part 50, Section 50.91 using the standards provided in Section 50.92. This analysis is provided below:

1. The proposed amendment will not involve a significant increase in the orobability or consecuences of an accident previousiv evaluated.

The proposed containment refueling integrity amendments do not affect the probability of a fuel handling accident, they only deal with the centainment systems.

The containment is provided for the purpose of mitigating the consequences of postulated accidents. For the fuel handling accident in containment, the licensing basis analyses, including the NRC safety .

I evaluation report transmitted February 2, 1982, assumed that containment was completely abrogated and all radioactive materials released from the containment refueling pool are assumed to be released to the outside atmosphere. The requested amendments to i Technical Specification 3.8.A.l.a modify the use of containment to )

mitigate the consequences of a fuel handling accident in containment, however, since instantaneous offsite release of all fuel handling Page 7 l

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, accident materials released to containment has already been considered, the probability and consequences of a loss of containment accident are not increased.

Therefore, the probability or consequences of an accident previously evaluated are not affected by any of the proposed amendments.

2. The proposed amendment will not create the possibility of a new or different kind of accident from any accident previous 1v analyzed.

The requested amendments to Technical Specification 3.8.A.l.a modify the use of containment to mitigate the consequences of a fuel handling accident in containment. There are no new failure modes or mechanisms associated with the proposed changes, nor do the proposed changes involve any modification of plant equipment or changes in plant operational limits. Previous analyses, including the NRC fuel handling accident safety evaluation for Prairie Island, have already assumed the containment is abrogated. The proposed license amendments may affect the release path for fission products released during a fuel handling accident in containment, but no new or different kind of accident will result.

Therefore, the possibility of a new or different kind of accident from any accident previously evaluated would not be created.

3. The proposed amendment will not involve a significant reduction in the marzin of safety The margin of safety as defined by the licensing bases fuel handling accident analyses is not reduced. The previous analyses are very conservative, assuming all radioactive material released from by the fuel handling accident is immediately released to the outside atmosphere, and bound any changes introduced by these requested amendments.

Technical Specification 3.8.A.1.a exists to minimize the consequences of a fuel handling accident in containment. However, with the current Technical Specification 3.8.A.l.a, there will still be releases due to the necessity to open the containment airlocks to evacuate personnel.

With implementation of this amendment, the ability of the closed airlocks to contain the accident releases may improve.

Some radioactive material could be released through containment penetrations that are open at the time of the accident. Since it is not likely that containment will be pressurized by a fuel handling accident, the releases are expected to be minimal. This amendment will maintain containment post-fuel handling accident offsite releases well within the limits of 10CFR100 and the current license basis releases.

Therefore, a significant reduction in the margin of safety would not be involved.

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Based on'the evaluation described above, and pursuant to 10 CFR Part 50, Section 50.91, Northern States Power Company has determined that operation of the Prairie Island Nuclear Generating Plant in accordance with the proposed license amendment request does not involve any significant hazards '

considerations as defined by NRC regulations in 10 CFR Part 50, Section 50.92.

Environmental Assessment -

Northern States Power Company has evaluated the proposed changes and determined that:

1. The changes do not involve a significant hazards consideration, or

2. The changes do not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or

3. The changes do not involve a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR Part 51 Section 51.22(c)(9).

Therefore, pursuant to 10 CFR Part 51 Section 51.22(b), an environmental assessment of the proposed changes is not required.

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1 Attachment 1 l Offsite Dose Assessment Following a Fuel Handling Accident The following assumptions and initial conditions are conservative bounds selected on the basis of worst case operating conditions and Regulatory Guide 1.25, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors":

1. The Fuel Handling Accident occurs 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after a shutdown, the minimum time permitted by Technical Specifications. Radioactivity decay is assumed during this time.

2. The low flow purge system is in operation during the Fuel Handling Accident. The exhaust dampers close but the purge system supply fan continues to force air into containment causing a small pressurization and the driving force for out flow.

3. The gap activity consisting of 10% of the total radioactive iodine in the rods at the time of the accident is assumed to be released from damaged rods. All dropped fuel assembly rods are assumed to suffer damage.

4. The iodine fission product activities are taken from a letter from J.A.

Usem (Westinghouse) to Bill Lax (NSP) dated September 13, 1993, Letter No. NSP-93-525. This source term is the total activity in the lead assembly with a burnup of 75,000 MWD /MTU. The radial peaking factor of 1.65 is applied since an average power level was assumed.

5. The overall effective decontamination factor of iodine in the water above the fuel is 100.

6. Radioactive material that escapes from the containment refueling pool mixes in containment and is released to the outside atmosphere exponentially (based on 6000 CFM) over a two hour time period through open airlocks and penetrations. This is conservative as no pressurization of containment is expected to drive the radioactive material out of containment.

7. No credit is taken for the atmospheric cleanup system in containment. This is conservative since the containment cleanup system (charcoal) ,

filters would be used to remove iodine.

8. The inhalation thyroid dose is determined using the assumptions of Section C.3.a of Regulatory Guide 1.25, except that updated ICRP-30 dose conversion factors are used.

9. No credit is taken for deposition of the plume on the ground or decay of isotopes in transit to the site boundary.

10. Atmospheric dispersion factors assume a ground level release. A conservative 0-2 hour site boundary X/Q of 6.5E 04 sec/m3 is taken from the Prairie Island USAR Appendix H.

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Attachment 2 Control Room Dose Assessment Following a Fuel Handling Accident The following assumptions and initial conditions are conservative bounds selected on the bas'is of worst case operating conditions and Regulatory Guide 1.25, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors":

1. The Fuel Handling Accident occurs 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after a shutdown, the minimum time permitted by Technical Specifications. Radioactivity decay is assumed during this time.

2. The gap activity consisting of 109 of the total radioactive iodine in the rods at the time of the accident is assumed to be released from damaged rods. All rods of the dropped fuel assembly are assumed to suffer damage.

3. The iodine fission product activities are taken from a letter from J. A.

Usem (Westinghouse) to Bill Lax (NSP) dated September 13, 1993, Letter No. NSP 93-525. This source term is the total activity in the lead assembly with a burnup of 75,000 MWD /MTU The radial peaking factor of 1.65 is applied since an average power level was assumed.

4. The overall effective decontamination factor of iodine in the water above the fuel is 100.

5. The entire inventory of radioactive material that escapes from the containment refueling pool is released to the outside atmosphere. This is conservative as no pressurization of containment is expected to drive the radioactive material out of containment.

6. No credit is taken for the atmospheric cleanup system in containment. This is conservative since the containment cleanup system (charcoal) filters would be used to remove iodine.

7. The inhalation thyroid dose is determined using the assumptions of Section C.3.a of Regulatory Guide 1.25, except that updated ICRP-30 dose conversion factors are used.

8. The Control Room Ventilation System fresh air purge factors, X/Q and iodine protection factors are taken from Fluor Daniel Calc. No. M-268-ZC-002, "SBVS Filter Efficiency Evaluation". This calculation assumes a control room iodine filter efficiency of 954, fresh air purges over a 2 day interval and unfiltered inleakage.

9. Atmospheric dispersion factors at the Control Room air intake is X/Q of 5.58E-03 soc /m3 .

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