ML20236Y492

Application for Amend to License DPR-46,accepting as-built Plant Configuration Involving Unreviewed Safety Question as Specified in Attachment 1

ML20236Y492
Person / Time
Site:	Cooper
Issue date:	08/06/1998
From:	Swailes J NEBRASKA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NLS980122, NUDOCS 9808120257
Download: ML20236Y492 (38)

Text

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H Nebraska Public Power District Nebraska's Energy Leader i

NLS980122 August 6,1998

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Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Gentlemen:

Subject:

Proposed License Amendment Secondary Containment Isolation Description and Refueling Accident Analysis Results Cooper Nuclear Station, NRC Docket 50-298, DPR-46

Reference:

1. NPPD Letter NLS970053, from Mr. P.D. Graham (NPPD) to USNRC dated March 31,1997,"USAR Rebaselining Project Description."

2. Cooper Nuclear Station Final Safety Analysis Report through Amendment 18

3. Letter from James R. Hall (USNRC) to Mr. Guy R. Horn (NPPD) dated January 27,1995," Cooper Nuclear Station - Amendment No.167 to Facility Operating License No. DPR-46 (TAC No. M89770)."

In accordance with the provisions specified in 10 CFR 50.59(c),10 CFR 50.90 and 10 CFR 50.4, the Nebraska Public Power District (District) requests that the NRC review and approve an amendment to accept an as-built plant configuration involving an unreviewed safety question as f

specified in Attachment 1. During the course of the Cooper Nuclear Station (CNS) Updated Safety Analysis Report (USAR) Rebaselining Project (Reference 1), the District discovered that USAR Sections V.3.2 and X.10.3.3.3 and Technical Specification Bases 3.2.D.2 state that the h4j Reactor Building Isolation and Control System isolates secondary containment sufficiently fast to prevent any release of fission products through the normal ventilation path following a Refueling Accident. The original plant design for which an Operating License was granted provides for isolation of the Secondary Containment such that during a Refueling Accident a limited amount of fission product release through the normal ventilation path may occur. Although the maximum theoretical release for this accident is higher than that originally evaluated by the NRC in Reference 2, there is no significant increase in the consequences of an accident as the off-site dose from this release is still well within the limits s;. -" led in 10 CFR 100. The District proposes to accept the original design and clarify the LoAR to reflect this design and resultant I

Cooper Nudear Station Po. Box 98/ Brownville, NE 68350098 1

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NLS980122 Page 2 of 4 analysis. This current plant configuration constitutes an unreviewed safety question requiring NRC approval in accordance with 10 CFR 50.59 since there is a potential limited amount of fission product release through the normal ventilation path during a Refueling Accident, which was not previously evaluated in the USAR.

The proposed license amendment is needed to accept plant configuration as-is and to clarify the USAR discussion of secondary containment isolation and the consequences of a release of fission products during a Refueling Accident to be consistent with the as-built plant. This proposed amendment will reflect the changes in analysis based on a 90 second closure time of a motor-operated Reactor Building i. solation damper, assuming a single failure of an air-operated Reactor Building isolation damper. Although the District is submitting this license amendment as a result ofits unreviewed safety question conclusion, the NRC should note that no actual change in the CNS Operating License is necessary. However, the District will update the USAR to clearly state that limited radioactive releases may occur during the Refueling Accident as a result of the 90 second closure time.

This proposed amendment does not alter the potential Control Room dose from the postulated Refueling Accident from that which has been previously submitted to the NRC. Amendment j

167 to the CNS Technical Specifications (Reference 3) allowed for increased flow capacity of the Control Room Emergency Filter System. The calculations used by the District in determining Control Room dose supporting Reference 3 assumed the 90 second closure time of the motor-

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operated damper.

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Attachment I contains a description of the unreviewed safety question requiring NRC approval and the attendant 10 CFR 50.92 evaluation. Attachment 2 contains anticipated marked-up USAR l

pages for informational purposes. The final USAR changes will be completed upon approval of the unreviewed safety question in accordance with 10 CFR 50.71(c). Attachment 3 contains USAR pages that were previously changed relative to the current plant configuration. Finally, provides a simplified diagram of the as-built plant configuration.

The discrepancies between the USAR description and as-built plant had been addressed in 1988 in 10 CFR 50.59 evaluations that focused on other aspects of this issue. It was determined at that l

time that an unreviewed safety question did not exist based on the District's and the industry's interpretation of the regulation. Ilowever, the District has reassessed its previous conclusions and factored in additional discrepancies recently discovered, and now concludes that the present configuration of the Reactor Building isolation dampers, when compared to the USAR and Technical Specification Bases descriptions, does constitute an unreviewed safety question.

The District has determined that because the unreviewed safety question involves secondary containment, the subject license amendment requires approval prior to unrestricted movement of fuel. Therefore, movement of fuel and heavy loads over irradiated fuel is presenth restricted.

This is an exigent issue for the District because the receipt of new fuel is expected mid-August of

NLS980122 l

Page 3 of 4 l

l 1998 in preparation for the Fall 1998 refueling outage. This new fuel requires inspection and i

subsequent movement into the spent fuel storage pool for temporary storage until the l

commencement of refueling activities. Therefore, the District respectfully requests exigent processing of this amendment, pursuant to 10 CFR 50.91(a)(6).

i This unreviewed safety question has been reviewed by the necessary Safety Review Committees.

l The District has concluded that the proposed changes do not involve a significant hazards consideration.

By copy of this letter and attachment the appropriate State of Nebraska official is being notified in accordance with 10 CFR 50.91(b)(1). Copies to the Region IV Office and the CNS Resident Inspector are also being sent in accordance with 10 CFR 50.4(b)(2).

Should you have any questions concerning this matter, please contact Mr. Brad Houston, Manager of Nuclear Licensing and Safety, at (402) 825-5819.

Sincerely, I

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W-l Jolun H. S vai s V"e Presi t of uclear Energy l

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Attachments cc: Regional Administrator w/ attachment USNRC - Region IV Senior Project Manager 10 copies w/ attachment USNRC - NRR Project Directorate IV-1 Senior Resident Inspector w/ attachment USNRC l

Environmental Health Division-Program Manager w/ attachment i

Nebraska Department of Health NPG Distribution w/o attachment

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l NLS980122 Page 4 of 4 I

I STATE OF NEBRASKA

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NEMAHA COUNTY

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1 John H. Swailes, being first duly sworn, deposes and says that he is an authorized representative f

of the Nebraska Public Power District, a public corporation and political subdivision of the State of Nebraska; that he is duly authorized to submit this corres o on behalf ofNebraska Public Power District; and that the statements contai aerein are true to the best of his 1

knowledge and belief.

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.. M r r Johni T

Swa les su ribed in i presence a id sworn to before me this day of To

,1998.

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Attachment I to NLS980122 Page 1 of 13 ATTACHMENT 1: PROPOSED LICENSE AMENDMENT SECONDARY CONTAINMENT ISOLATION DESCRIPTION AND REFUELING ACCIDENT ANALYSIS RESULTS

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COOPER NUCLEAR STATION 1

NRC DOCKET NO. 50-298, LICENSE DPR-46

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' INTRODUCTION The Nebraska Public Power District (District) requests that the NRC approve an existing plant design and accident analysis which involves an unreviewed safety question as described below.

The purpose of this proposed amendment is to obtain approval of the as-built configuration and i

the revised Refueling Accident analysis of the Secondary Containment and Reactor Building Isolation and Control System. The discussion of the Secondary Containment and the Reactor Building Isolation and Control System in the Cooper Nuclear Station (CNS) Updated Safety

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Analysis Report (USAR) may then be clarified to reflect the revised analysis for the Refueling.

Accident based on the as-built design. The as-built configuration consists of two trains of L

isolation dampers in the Reactor Building normal ventilation exhaust path, each train having an j

air-operated and motor-operated damper in series. The current USAR is confusing in that it

' indicates that the postulated Refueling Accident results in limited fission product releases L

through the normal ventilation exhaust path, and at the same time also indicates that no releases 3

l-would occur.

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The proposed USAR changes are included in Attachment 2 for your information. The final li USAR changes will be processed in accordance with 10 CFR 50.71(e) pending approval of this proposed license amendment. Attachment 3 includes USAR changes which were previously made, also provided for your information.

The District respectfully requests exigent processing of this amendment, pursuant to 10 CFR f

50.91(a)(6), in support of the upcoming refueling outage. The basis for this request is to avoid impacting the refueling outage, scheduled to begin October 2,1998. The District has determined that because the unreviewed safety question involves secondary containment, the subject license amendment requires approval prior to unrestricted movement of fuel. Therefore, movement of fuel and heavy loads over irradiated fuel is presently restricted. This is an exigent issue for the Distri the receipt of new fuel is expected mid-August of 1998 in preparation for the Fall 1998.

Eg outage. This new fuel requires inspection and subsequent movement into the spent fuel storap pool for temporary storage until the commencement of refueling activities.

For a Refueling Accident, the original safety analysis as described in USAR Sections V.3.2 and X.10.3.3.3 assumed fast closure of the Secondary Containment (Reactor Building) isolation dampers and concluded postulated fission product releases would be less than 1 Rem (off-site l

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- Attachment I J

to NLS980122 Page 2 of 13 dose to the thyroid and to whole body). The CNS Preliminary Safety Analysis Report (PSAR) was submitted to the Atomic Energy Commission (AEC) in July of 1967 with indication that no releases through the normal discharge path would result from the postulated Refueling Accident.

This was largely due to an assumed fast closure time (three seconds) for the reactor building isolation dampers combined with a six_ second hold up time inherent in the design of the ventilation system duct work.- Amendment 3 to the PSAR (February 1968) discussed the Reactor Building ventilation system six second hold up feature which ensured sufficient time for closure of the Secondary Containment isolation dampers before contaminated air would reach that location. A 1968 revision of a Reactor Building Heating, Ventilating, and Air Conditioning drawing also illustrated two air-operated dampers in series. Sometime later the system design was changed to provide diverse isolation and Revision 2 to the drawing illustrated the current configuration of one air-operated damper in series with one motor-operated damper (dated 1970).

The later configuration is what is installed in the plant. A simplified diagram of this configuration is included in Attachment 4.

I The CNS Final Safety Analysis Report (FSAR) was submitted in 1971 to the AEC, and this submittal briefly discussed the configuration of the dampers (air and motor-operated dampers in

' series). In 1972, Amendment 9 to the FSAR was submitted to the AEC in response to FSAR Question 9.14, where the AEC questioned the total time to switch from a normal Reactor Building lineup to standby gas treatment system operation upon detection of a high radiation signal. The District responded by citing the six second response time to achieve isolation damper closure. This six second time would correlate to the air-operated damper. However,if appropriate single failure criteria were applied to that accident, the District should have assumed that the air-operated damper would fail and provided the closure time for the motor-operated valve (60 to 90 seconds).

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To further complicate the matter, the full power Operating License issued to CNS by the NRC cn January 18,1974, was accompanied by a Safety Evaluation Report (SER), issued Febrary 14, 1973. Section 15.3," Refueling Accident," of the SER states that, for the Refueling Accident, "The airborne fission products within the building are assumed to pass through the standby gas l

treatment system '... " which implies that no fission products are released through the normal

- discharge path. The SER also implies that the calculated radiological doses that might result from any of the potential design basis accidents (of which the Refueling Ace' dent is one) are well

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within the guidelines given in 10 CFR 100. Notwithstanding this license amendment request and

unreviewed safety question, the radiological doses will remain within this NRC threshold.

In 1988, it was recognized that the as-built configuration of the plant was not consistent with the USAR description for Secondary Containment. Corrective actions resulted in the revision of the USAR to reflect the existence of a motor-operated damper which has a 60 second closure time.

I As a result of this configuration, the revised Refueling Accident analysis reflected the potential for limited releases from the Reactor Building through the normal discharge path during the time l

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4 Attachment I to NLS980122 Page 3 of 13

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it takes for the dampers to achieve full closure. The revised analysis also assumed the release point equivalent to the turbine building roof (using 40 meter release values contained in USAR Table XIV-8-3), and assumed constant flow through the drmpers until they were fully closed.

Once the dampers isolated, releases would be from the Standby Gas Treatment (SGT) System.

This resulted in a change in the consequences of this accident from less than 1 Rem maximum

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off-site dose to the thyroid to 1.8 Rem, still a small fraction of 10 CFR 100, but an increase nonetheless. The dose to the whole body remained at less than 1 Rem. The change in resultant i

dose was screened for whether it constituted an unreviewed safety question. It was determined the change did not involve an unreviewed safety question because the dose consequences remained well within the limits imposed by 10 CFR 100. The USAR was revised pursuant to 10 CFR 50.71(e) to incorporate these changes without requiring prior NRC review and approval.

However, this 10 CFR 50.59 screen did not appear to address the differences in the air-operated damper assumed to be in place in the NRC SER versus the actually installed motor-operated damper, and the resultant impact on no release versus a limited release.

In 1993, the Refueling Accident analysis was again revised to accommodate a surveillance procedure change reflecting a more conservative 90 second closure time of the dampers, based on uncertainties in stroke timing. The revised analysis resulted in a maximum off-site dose of 2.7 Rem to the thyroid, still a small fraction (less than 1%) of 10 CFR 100. The dose to whole j

body remained at less than 1 Rem. The change was screened under 10 CFR 50.59, using the guidance for safety evaluations at that time, and was again determined not to involve an unreviewed safety question since the consequences still remained well within the limits imposed by 10 CFR 100. The USAR was updated to incorporate the analysis results and did not receive

. prior NRC review and approval. Attachment 3 contains the USAR changes for informational purposes.-

As part of the USAR Rebaselining Project"1, a further review of these changes to the Refueling

' Accident Analysis in the USAR in May 1998 led to the discovery ofinconsistencies with respect j

to this analysis in the Secondary Containment safety design bases discussion in the USAR (Section V.3.2), and the description of the heating, ventilation, and air-conditioning system in p

Secondary Containment (Section X 10.3.3.3). In July 1998, the District concluded that these changes htvolved an unreviewed safety question requiring NRC approval. Consistent with the j

guidance provided in the District's recently revised procedure governing changes pursuant to 10 CFR 50.59, this conclusion is based upon the fact that the original design basis states no release via the normal ventilation discharge path. However, the revised analysis for the as-built

plant design resulted in limited releases through the normal discharge path, which also resulted in

. a potential increase in the postulated radiological consequences of an accident previously evaluated in the USAR. The consequences are still well within the limits of 10 CFR 100.

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' Reference NPPD Letter NLS970053, from Mr. P. D. Graham (NPPD) to USNRC dated March 31,1997,"USAR Rebaselining Project Description."

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Attachment I to NLS980122 Page 4 of 13 This proposed amendment does not alter the potential Control Room dose from the postulated Refueling Accident from that which has been previously submitted to the NRC. Amendment l

167t21 to the CNS Technical Specifications was granted January 27,1995, which allowed for l

increased flow capacity of the Control Room Emergency Filter System. The calculations used by l

the District in determining Control Room dose supporting the District's request for Amendment 167 assumed the 90 second closure time of the motor-operated damper!'l.

The analys ; e also consistent with the Improved Technical Specifications (ITS) Bases for the trip settings for the Reactor Building Ventilation Exhaust Plenum radiation monitors (ITS is scheduled for implementation at CNS in August 1998). The allowable value in ITS for the trip setting on high radiation in the Reactor Building is less than or equal to 49 mR/hr (ITS Table 3.3.6.2-1). The basis for this setting, as stated in ITS Bases Section B 3.3.6.2, is to provide indication of possible gross failure of the fuel cladding, due to a break in the reactor coolant pressure boundary or Refueling Accident, and initiate appropriate actions. When high radiation is detected, Secondary Containment isolation and actuation of the SGT System are initiated to 1

limit the postulated release of fission products assumed in the USAR safety analyses. The Bases j

also state that the allowable values are chosen to promptly detect gross failure of the fuel

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cladding. The trip settings, operability requirements, and surveillance requirements are l

established such that postulated off-site dose limits are not exceeded.

2.0 DISCUSSION 2.1 System Description and Safety Basis The Secondary Containment System is designed to provide secondary containment for the postulated Loss of Coolant Accident (LOCA) and primary containment for the postulated Refueling Accident. The CNS USAR states that the principle safety objective of the Secondary Containment is to limit the release of radioactivity to the environs after an accident so that the resulting exposures are kept to a practical minimum and are within the guideline values given in 10 CFR 20 and 10 CFR 100. In the event of a release of radioactivity to the Reactor Building atmosphere, the Secondary Containment contains the necessary reliable and redundant components / subsystems to isolate, contain, and assure controlled filtered elevated release of the Reactor Building atmosphere.

2 Letter from James R. Hall (USNRC) to Guy R. Horn (NPPD), " Cooper Nuclear Station Amendment No.167 to Facility Operating License No. DPR-46 (TAC No. M89770)."

' NPPD Letter NLS940002 from G.R. llorn (NPPD) to USNRC dated July 20,1994,

" Control Room Emergency Filter System Commitments."

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Attachment I to NLS980122 Page 5 of 13 Key e!ements of the safety design bases for Secondary Containment are described below and may be found in CNS USAR Section V.3.2:

The secondary containment system is designed with sufficient redundancy so that no single active system component failure can prevent the system from achieving its safety objective.

The secondary containment is designed to limit the ground level release to the environs of airbome radioactive materials so that off-site doses from a design basis fuel handling or loss of coolant accident (LOCA) will be below the values stated in 10 CFR 100.

The reactor building isolation and control system is designed to isolate the reactor building sufficiently fast to prevent fission products from the postulated Refueling Accident from being released to the environs through 4

the normal discharge path.

The Secondary Containment utilizes the following four different features to mitigate the consequences of a postulated LOCA and the Refueling Accident (USAR V.3.3.1):

A negative pressure barrier relative to the building exterior which minimizes the ground level release of fission products by exfiltration.

A low leakage containment volume which provides a hold up time for fission product decay prior to release.

The removal of particulate and iodines by filtration prior to release through the SGT System.

The exhausting of the Secondary Containment atmosphere through the i

Elevated Release Point which aids in the dispersion of the effluent by atmospheric diffusion.

In the event high radiation is detected in the Secondary Containment normal ventilation exhaust plenum (due to a postulated Refueling Accident), the Reactor Building Isolation and Control System initiates isolation of the reactor building normal ventilation system, which includes tripping the supply and exhaust fans, and starts the SGT System.

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Attachment I to NLS980122 Page 6 of13 The District has determined that the revised analysis is acceptable and thus the as-built plant design is also acceptable. Therefore, the licensing basis requires reconciliation with the existing plant design. Si ice the District determined this involves an unreviewed safety question (as discussed previously), NRC approval of the as-built design is also required.

2.2 Exigent Circumstances In accordance with 10 CFR 50.91(a)(6), the District requests that this amendment be treated as an exigent amendment. This resulted from the fact that the District recently determined the two previous USAR changes, which were processed in 1988 and again in 1993 pursuant to 10 CFR 50.71(e), involved an unreviewed safety question with respect to the revised Refueling Accident analysis. However, at the time the USAR changes were made,it was determined that an unreviewed safe' question did not exist. The adequacy j

of the unreviewed safety question evaluations for these USAR changes were called into question by the USAR Rebaselining Project in May 1998. A reevaluation of the USAR

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changes pursuant to 10 CFR 50.59 was completed in July 1998, which determined that

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NRC review and approval should have been sought prior to implementing the changes into the CNS USAR. A License Amendment has been immediately prepared and is being submitted within three weeks of the District's conclusions regarding this matter.

l It is also noteworthy that the Technical Specification Bases Section 3.2.D.2 states " Trip settings ofless than 100 millirem per hour (mR/hr) for the monitors in the ventilation i

exhaust ducts are based upon initiating normal ventilation isolation and Standby Gas Treatment System operation so that none of the activity released during the refueling accident leaves the Reactor Building via the normal ventilation path but rather all the activity is processed by the Standby Gas Treatment System." The District has examined options to identify compensatory measures which could be taken to allow unrestricted i

fuel movement to occur on schedule and still remain within the license conditions in terms of dose consequences. However, these factors do not reconcile design basis of the Secondary Containment System that no fission products will be released via the normal discharge path, and therefore a license amendment prior to unrestricted fuel movement l

must be obtained. The District has therefore restricted movement of fuel or heavy loads

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over irradiated fuel.

Receipt of new fuel is expected mid-August of 1998 in preparation for the Fall 1998 refueling outage, scheduled to begin October 2,1998 Upon receipt, this new fuel requires inspection and subsequent movement into the spent fuel storage pool for temporary storage until the commencement of refueling activities. Delay in the receipt

1 Attachment I to NLS980122 Page 7 of13 inspection and movement of this fuel could have significant impacts on the District.

Therefore, the District respectfully requests exigent processing of this amendment, pursuant to 10 CFR 50.91(a)(6), to avoid delaying the upcoming refueling outage.

3.0 DESCRIPTION

OF CllANGES This proposed license amendment requests NRC review and approval of a de-facto change in the facility. The change, as described previously, is the as-built configuration of an air-operated damper in series with a motor-operated damper in the two normal ventilation exhaust ducts of Secondary Containment (Reactor Building). This change impacts the Refueling Accident analysis in that the potential for limited releases through the normal discharge path exists. This also results in an increase in the radiological dose, however this dose remains a small fraction (less than 1%) of the limits specified in 10 CFR 100. Therefore, the District proposes to accept this as-built plant configuration and associated analysis. The impact of this as-built configuration on Control Room dose was previously submitted to the NRC in support of Amendment 167. contains a discussion of the proposed changes and marked-up copies of the proposed USAR changes and is provided for your information. The final USAR changes will be processed in accordance with 10 CFR 50.71(e).

3.1 Justification As discussed previously, the District proposes to accept the as-built configuration of the air-operated damper in series with the motor-operated damper and the associated Refueling Accident analysis. The as-built design provides for sufficient redundancy and diversity so that no single active system component failure can prevent the system from achieving its safety objective. In addition, defense in depth is maintained because the supply and exhaust fans are designed to trip on an isolation signal, however in the analysis they are assumed to continue running, providing a motive force for air movement outside of Secondary Containment (Reactor Building). This configuration, assuming a 90 second closure time, does not alter Control Room dose which was previously submitted by the District in support of obtaining Amendment 167, as the submittals assumed the single failure of an air-operated damper and 90 second cicrure time for the motor-l operated damper.

The Refueling Accident analysis is also sufficiently conservative in its assumptions regarding the potential activity released from the fuel in a postulated Refueling Accident.

The original analysis utilized General Electric (GE) 7x7 fuel bundle design. USAR Section XIV.6.4.8 identifies that the radiological exposures for the 7x7 fuel are well below the guidelines set forth in 10 CFR 100. Ilowever, since that time Cooper Nuclear Station (CNS) has utilized GE's 8x8 designs and 9x9 Lead Test Assembly (LTA; also I

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Attachment I to NLS980122 Page 8 of 13 called Lead Use Assembly) designs. Analyses have shown that, for the postulated Refueling Accident, the fission product activity released is less for the 8x8 and the 9x9 LTA designs than that for the 7x7 design. Thus, since the radiological exposures for the 7x7 fuel are well below the regulatory threshold speciDed in 10 CFR 100, it can be concluded that the consequences of this accident for 8x8 and the 9x9 LTA c'esigns will also be well below these guidelines.

A Probabalistic Risk Assessment analysis was performed to evaluate the reliability of the Secondary Containment isolation function in response to a Refueling Accident. The evaluation compared the reliability of the damper conGguration of two air-operated dampers in series to the as-built conHguration. The analysis concluded that the as-built design of an air-operated damper and motor-operated damper in series provides for increased reliability of the isolation function due to the diversity of damper types. This is based on the calculated probability of the failure to isolate for each configuration and factoring in a common cause failure estimation for the dual air-operated damper con 0guration. The probability of the failure to isolate for the two air-operated dampers in series was calculated to be 6.2E-4, and the probability of the failure to isolate for the air-operated damper in series with the motor-operated damper was calculated to be 1.8E-5.

Since the radiological consequences are a small fraction of 10 CFR 100 (less than 1%),

and the as-built design provides for adequate redundancy, diversity, and improved reliability over the connguration of two air-operated dampers in series, the District concludes that the as-built design is acceptable.

4.0 NO SIGNIFICANT HAZARDS CONSIDERATION EVALUATION 10 CFR 50.91(a)(1) requires that licensee requests for operating license amendments be accompanied by an evaluation of signincant hazards posed by the issuance of the amendment.

This evaluation is to be perfonned with respect to the criteria given in 10 CFR. 50.92(c). The I

following analysis meets those requirements.

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Attachment I to NLS980122 Page 9 of 13 Evaluation of this Amendment with Respect to 10 CFR 50.92 The enclosed proposed license amendment for the as-built design of the Secondary Containment (Reactor Building) isolation dampers isjudged to involve no significant hazards based on the following:

1. Does not involve a significant increase in the probability or consequences ofan accident previously evaluated The existing plant design does not involve a significant increase in the probability of an accident previously evaluated in the Updated Safety Analysis Report (USAR). The current configuration does not affect the performance and reliability of the Secondary Containment and the Reactor Building Isolation and Control System or any system interface in a way that could lead to an accident occurring. The current configuration and analysis do not affect any accident precursors or initiators, and therefore, does not increase the probability of an accident.

The present plant configuration also does not involve a significant increase in the consequences of an accident previous?y evaluated in the USAR. The current design will require a clarification to the Secondary Containment safety design basis as described in the USAR to reflect the as-built configuration and analysis of the plant by stating that the R.eactor Building Isolation and Control System is designed to limit the release of fission products through the normal ventilation discharge path during a postulated Refueling Accident. The original analysis determined that the consequences of the Refueling Accident were significantly less than 1 Rem to the thyroid and whole body (maximum off-site dose). When l

this analysis was revised to account for the 90 second motor-operated damper closure time, the calculated whole body off-site dose increased, but was still less than 1 Rem; the calculated off-site dose to the thyroid, however, increased to 17 Rem. While this change in the analysis represents an order of magnitude increase in consequences (thyroid dose increase i

from 17 millirem to 2.7 Rem), the actual increase is minimal because this increase in l

l consequences is still less than 1 per cent (1%) of the limits specified in 10 CFR 100. Thus 1

the consequences still remain well within the regulatory threshold specified in 10 CFR 100 and thus pose no undue hazard to the health and safety of the public. This proposed j

l amendment does not alter the Control Room dose from that which was submitted to the NRC l

in support of Amendment 167.

2. Does not create the possibilityfor a new or different kind ofaccidentfrom any accident previously evaluated This proposed license amendment is administrative in nature in that it reflects the effects of a revised analysis for the Refueling Accident, which is an accident previously analyzed as a Design Basis Accident (DBA) in the SAR, based on the present configuration of the plant.

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Attachment I to NLS980122 Page 10 of 13 The current configuration does not create the possibility of a new or different kind of accident i

from any accident previously evaluated in the USAR. The proposed license amendment does i

not introduce any new equipment or hardware changes, nor does it require existing equipment or systems to perform a different type of function than they are presently designed to perform.

The as-built configuration does not introduce any new mode of plant operation, thus there are l

no new accident failure paths created. The as-built configuration does not affect any accident precursors or initiators and does not create the possibility of a new or different kind ofaccident.

3. Does not create a sigmficant reduc. ion in the margin ofsafety The present plant configuration does not involve a significant reduction in a margin of safety.

Technical Specification Bases section 3.2.D.2, Reactor Building Isolation and Standby Gas Treatment (SGT) Initiation, states that the trip settings for the Reactor Building exhaust l

plenum radiation monitors are based upon initiating normal ventilation system isolation and SGT System operation so that none of the activity released during the refueling accident leaves the Reactor Building via the normal ventilation path, but rather all the activity is processed by the SGT System. This basis statement remains true unless there is a single failure of the air-operated Secondary Containment isolation damper. Under single failure conditions there would be the potential for a limited release through the normal ventilation system prior to complete isolation of the secondary containment and initiation of the SGT System. The significance of this change is minimal, as Technical Specification requirements to isolate Secondary Containment are still met. The overall function of the Secondary Containment and Reactor Building Isolation and Control System, in conjunction with other accident mitigation systems, is to limit fission product release during and following postulated DBAs. liigh radiation in the Secondary Containment exhaust is an indication of possible gross failure of the fuel cladding, possibly due to a Refueling Accident. The trip settings for the Reactor Building (Secondary Containment) radiation monitors are such that initiation of secondary containment isolation and SGT would still occur in sufficient time (within 90 seconds of detection) to maintain postulated off-site releases well within the limits of 10 CFR 100. As stated previously, the effects of the 90 second motor-operated damper closure time on Control Room dose have already been taken into consideration in the District's submittals supporting Amendment 167.

5.0 ENVIRONMENTAL IMPACT EVALUATION 10 CFR 51.22(c)(9) provides criteria for, and identification of, licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration, (2) result in a significant change in the types or significant

Attachment I to NLS980122 Page 11 of 13 l

increase in the amount of any effluents that may be released off-site, or (3) result in an increase in individual or cumulative occupational radiation exposure. The District has reviewed the l

proposed license amendment and concludes that it meets the eligibility criteria for categorical l

exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(c), no environmental impact statement or environmental assessment needs to be prepared in connection with issuance of the proposed license changes. The basis for this determination is as follows:

1.

The proposed license amendment does not involve significant hazards as described previously in the No Significant Hazards Consideration Evaluation.

l 2.

As discussed in the No Significant Hazards Consideration Evaluation, this proposed amendment to accept the es-built configuration of the plant does not result in a significant increase in radiological doses for any Design Basis Accident. This proposed license amendment does not result in a significant change in the types or significant increase in the amounts of any effluents that may be released off-site. The proposed license amendment does not introduce any new equipment, nor does it require any existing equipment or systems to perform a different type of function than they are presently designed to perform. The i

District has concluded that there will not be a significant increase in the types or amounts of any effluents that may be released off-site and these rhanges do not j

involve irreversible environmental consequences beyond those already associated with normal operation, j

3.

The proposed license amendment involves a change in the closure time for the motor-operated dampers, resulting in a potential increased off site radiological dose in the event of a Refueling Accident accompanied by a single failure of an air-operated damper (in series with the motor-operated damper). The proposed license amendment does not inct nse individual or cumulative occupational radiation exposure beyond that already associated with normal operation.

6.0 CONCLUSION

The District has evaluated the proposed changes to the USAR to clarify the discussion of the Secondary Containment and the Reactor Building Isolation and Control System to reflect the revised analysis for the Refueling Accident against the criteria given in 10 CFR 50.92(c) in accordance with the requirements of 10 CFR 50.91(a)(1). This evaluation has determined that the proposed changes will not: 1) involve a significant increase in the probability or consequences of an accident previously evaluated; 2) create the possibility for a new or different kind of accident from any previously evaluated; or 3) create a significant reduction in the margin of safety. Therefore, for the reasons detailiid above, the District requests NRC approval of this proposed amendment.

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L Attachment I to NLS980122.

Page 12 of13

7.0 REFERENCES

. 7.1 10 CFR Part 20 7.2 10 CFR Part 50, Sections 50.4,50.59,50.71(c),50.90,50.91,50.92 7.3 10 CFR 51.22

~i p

7.4 10 CFR 100 l

7.5 CNS Updated Safety Analysis Report (USAR) Chapters I, V, X, and XIV 7.6 CNS Calculation NEDC 88-171 Rev.1," Refueling Accident (7x7 bundles)-

Radiological Effects - 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> dose."

l l

7.7 CNS Calculation NEDC 88-175 Rev.1," Fuel Handling Accident Dose Calculation."

7,8 CNS Calculation NEDC 94-071 Rev. 5," Control Room Operator Dose Due to Inleakage to Control Room."

l 7.9 CNS Final Safety Analysis Report (FSAR) through Amendment 18 7.10 NPPD Letter NLS970053, from Mr. P. D. Graham (NPPD) to USNRC dated l

March 31,1997,"USAR Rebaselining Project Description."

7.11 GESTAR II, NEDE-24011-P-A-11-US 7.12 Burns and Roe Drawing 2020, Rev. N39, " Flow Diagram, Reactor Building, Heating & Ventilating."

7.13 NPPD Letter NLS940002 from G. R. Horn (NPPD) to USNRC dated July 20, l

1994," Control Room Emergency Filter System Commitments."

7.14 NPPD Letter NLS940006 from G. R. Horn (NPPD) to USNRC dated July 26, 1994, " Proposed Change No.135 to Technical Specifications Control Room L

Emergency Filter System."

7.15 NPPD Letter NLS940143 from G. R. Horn (NPPD) to USNRC dated I

December 27,1994," Revision to Proposed Change No.135 to Technical Specifications Control Room Emergency Filter System."

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L-_-__-----______--__

Attachment I to NLS980122 Page 13 of13 7,16 Letter from James R. Hall (USNRC) to Guy R. Horn (NPPD) dated January 27, 1995,"Cocper Nuclear Station - Amendment 167 to Facility Operating License No. DPR-46 (TAC No M89770)."

7.17 CNS Operating License No. DPR-46," Nebraska Public Power District, Docket No. 50-298 (Cooper Nuclear Station) Facility Operating License," Issued January 18,1974.

7.18 Section 15,3 of" Safety Evaluation by the Directorate of Licensing, U.S. Atomic Energy Commission, In the Matter of Nebraska Public Power District, Cooper Nuclear Station, Nemaha County, Nebraska, Docket No. 50-298," Issued February 14,1973.

i

ATTACHMENT 2 to NLS980122 COOPER NUCLEAR STATION NRC DOCKET NO. 50-298, LICENSE DPR-46 PROPOSED USAR CHANGES FOR INFORMATION ONLY AFFECTED PAGES:

I-8-2 V-3-1 X-10-6 XIV-9-1 DESCRIPTION OF PROPOSED CHANGES The following provides a summary of the proposed changes to the USAR and ITS. Marked-up pages of the USAR changes are also included for information. Changes to Technical Specifications Bases Section 3.2.D.2 will not be made since it is anticipated that Improved Technical Specifications (ITS) shall become effective prior to issuance of this license amendment. Clarifications to ITS Bases which are required as a result of this license mnendment shall be processed in accordance with 10 CFR 50.59. Final changes to the USAR shall be processed in accordance with 10 CFR 50.7)(e).

1.

USAR Volume I, Table 1-8-1, " Summary ofMaximum Off-Site Effects ofDesign Basis Accidents, "page 1-8-2 Note 5 to this table currently states," Based on an effective release height of 40 meters for the first minute,.. "

The proposed revision to this note states," Based on an effective release height of 40 meters for the first 90 seconds... "

2.

USAR Volume 11,Section V.3.2, Secondary Containment System Safety Design Basis, page V-3-1.

Safety Design Basis No. 6 in this section states that,"The reactor building isolation and control system is designed to isolate the reactor building sufficiently fast enough to prevent fission products from the postulated Refueling Accident from being released to the environs through the normal discharge path."

The proposed revision to this criterion states, "The reactor building isolation and control system is designed to isolate the reactor building sufficiently fast to limit fission product release through the normal discharge path during the postulated e

l l

Refueling Accident. Any potential off-site doses from this release will be well within the values stated in 10 CFR 100."

3.

USAR Volume IV, Section X10.3.3.3, Secondary Containment, page X-10 6.

This section states within the discussion contained in paragraph two that,"This I

long duct provides hold up time to ensure no release ofcontamination to atmosphere (which may be detected by the Radiation Monitor) during the time required for the signal from the Radiation Monitor to actuate the exhaust valves in full closure position."

The proposed revision states,"This long duct provides hold-up time to limit a fission product release to the atmosphere through the normal discharge path I

(which may be detected by the Radiation Monitor) during the time required for the l

signal from the Radiation Monitor to actuate the exhaust valves in full closure I

position."

4.

USAR Volume V,Section XIV.9, Dose Sensitivity Evaluation Using Assumptions ofthe AEC/DRL, page XIV-9-1.

It is proposed to add a note to this section to state that,"The information presented in this section is for historical purposes only and is not updated."

i 5.

ITS Bases, Section B 3.6.4.2, pages B 3.6-71 and B 3.6-72 (scheduled implementation date is August 15,1998)

The Background section currently states that "... fission products... are maintained within the secondary containment boundary" and that automatic Secondary Containment Isolation Valves (SCIVs, which are the air-operated dampers in series with the motor-operated dampers)"... close on a Secondary Containment isolation signal to establish a boundary for untreated radioactive material within Secondary Containment following a DBA [ Design Basis Accident] or other accidents."

With respect to operability of the SCIVs and the LOCA and Refueling Accident, the Applicable Safety Analysis section currently states,"The secondary containment performs no active function in response to either of these limiting events, but the boundary established by the SCIVs is required to ensure that leakage from the primary containment is processed by the Standby Gas Treatment System (SGT) before being released to the environment."

The proposed revisions to the Background and Applicable Safety Analysis sections would reflect the existing plant configuration and Refueling Accident analysis results which conclude that a limited fission product release through the normal discharge path occurs.

I

-s USAR TABLE I-8-1

SUMMARY

OF MAXIMUM OFF-SITE EFFECTS OF DESIGN BASIS ACCIDENTS'"

t Maximum Off-Site Dose.(rems)

Desian Manis Accident Deen-Dose Eauivalent Thyroid Control Rod Drop Accident'28 83' 4.0 x 10*

4. 4 x 10*'

Loss of Coolant Accidenti2# (4)

,2.3 x 10-8 2.0 x 10'*

Refueling Accident'82 "8 4.1 x 10-2 2.7

. Main Steam Line Break AccidentW

1. 9 x 10-'

7.0 x 10-1 l

-(1)

Maximum Off-Site exposures are for 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> exposures at the site boundary.

1 l

(2)

Based on. analysis for original'7x7 fuel.

(3) Based on an effective release height'of 30' meters for Control Rod Drop Accident.

~(4)

Based on an' effective release height of 100 meters for Loss of Coolant Accident.

l (5). Based on an effective release height of 40 meters for the first meende 90gseconds of. unfiltered release, followed by an effective release.leight of 100 meters for the duration of the accident for the Refueling Accident, l

j, (6)

Based on an effective release height of 30 meters for the Main Steam Line Break Accident.

4

)

I-8-2 07/22/96 i

i

-J__---__-.--_--__-____

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USAR 3.0 SECONDARY CONTAINMENT SYSTEM l

3.1 Safetv obiective The safety objective of the secondary containment system in conjunction with other engineered safeguards and nuclear safety systems is to l

limit the release to the environs of radioactive materials so that off-site doses from a postulated design basis accident will be below the values of 10CFR100.

i 3.2 Safetv Desion Basis The safety design bases of the secondary containment system are as i

follows:

1.

The secondary containment system shall provide secondary containment when the primary containment is operable and primary containment when the primary containment is open.

2.

The secondary containment system is designed with sufficient i

redundancy so that no single active system component failure can prevent the l

system from achieving its safety objective.

3.

The secondary containment system is designed in accordance with Class I design criteria.

The exception to the Seismic Class IS design criteria is that piping which penetrates Secondary Containment need not meet Seismic Class IS requirements at the SC penetration. Piping which is credited for mitigation of accidents, is designed to IS requirements to the IS anchor location / isolation valve inside Secondary Containment (e.g., Main Steam Lines).

4.

The secondary containment is designed to limit the ground level release to the environs of airborne radioactive materials so that off-site doses from a design basis fuel handling or loss of coolant accident (LOCA) will be below the values stated in 10CFR100.

5.

The secondary containment system is designed to be sufficiently leaktight to allow the Standby Gas Treatment (SGT) system to reduce the reactor i

building pressure to a minimum subatmospheric pressure of 0.25 inches of water (under neutral wind conditions) when the SGT system fans are exhausting reactor building atmosphere at a rate of 100% per day of the reactor building free L

volume.

p b

. ova.

- M

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my cu 9_"' hb " > "_ _ h " M"a p"y v"m" ","yu.

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-m m um m vya - u m m.m. - um vc emu.

METWO 6M 7The.'_ react 6r'bQilding fisolation5and? control? system Mis Tdesignsd to isolate t the e reactor? building, sufficientiv.. fastmto.111mith fission aproduct

, release G throught the e normal" discharge: pathe duringsthe postulated t refueling accident'. Any potential off-site Ldoses from this release will berwell within: the

~

~ ~

'u

~~

yalues;statedjin;10lCFR(100; 7.

The secondary containment system is provided with means to conduct periodic tests to verify system performance.

3.3 nupriot ion 3.3.1 General

~

The secondary containment system consists of four subsystems. These l

subsystems are the reactor building, the reactor building isolation and control system, the SGT system and the Elevated Release Point (ERP).

The SGT system is supported in its function by the Sump Z system which removes condensation generated in the SGT discharge line'during SGT system operation (see X-14.0 for details on Sump Z). The secondary containment system surrounds the primary containment system and is designed to provide secondary containment for the postulated LOCA.. The secondary containment system also surrounds the refueling facilities and is designed to provide primary containment for the postulated refueling accident.

The secondary containment system utiliza four different features to mitigate the consequences of a postulated LOCA (pipa break inside the drywell) and the refueling accident (fuel handling accident).

The first feature is a negative pressure barrier which minimizes the ground level release of fission products by exfiltration. The second feature is a low leakage containment volume which provides a hold-up time for fission product decay prior to release. The l

l V-3-1 03/11/98 l

USAR Four fan-coil units are available, with all normally operating.

Cooling water is supplied by the Reactor Equipment Cooling system to the unit coils.

Each fan-coil motor is controlled manually from the main control room with pilot light indication of which units are operating. Temperature of the fan discharge is recorded and is annunciated when it exceeds 5" above the maximum i

setting.

The drywell and the pressure suppression chamber are connected, through suitable valving, to the Reactor Building general exhaust system and also l

to the Standby Gas Treatment (SGT) System.

This provides purge and clean-up j

capability. The primary containment is also connected to the Nitrogen Inerting System which can provide an atmosphere high in nitrogen content and low in oxygen content.

Refer to Section V-2,

" Primary Containment System," for details.

l 10.3.3.3 Secondary containment i

The secondary containment area (Reactor Building) has supply and l

exhaust ventilating systems. The supply system furnishes filtered and tempered i,

100% outdoor air to all floors of the building through ductwork.

Heating and ventilating units each consisting of an automatic filter, heating coil, air i

washer and duplex fans will be furnished, as described in Section X-10.3.4.2.

l The exhaust air is induced from all but the area below grade level, 1

through the open hatches and stairwells up to the operating floor. All of this air, plus the ventilation air directly supplied to the operating floor, flows across the pools to multiple openings in the sides of the pools just above water I

level. This pattern provides an air blanket at the pool and prevents air flow from this area flowing to other areas of lower contamination potential. The air flows through embedded ducts in the concrete wall of the pools and then via exposed ductwork to a common plenum connected to the two exhaust fans, each of 100% capacity. The air is then exhausted to the atmosphere through a long duct run. This long duct provides hold-up time to incur; n: r:1 ;;; cf ;;nt:mination Ot;;;N;;; $!IU$$2MO[j:R[BjMREL1dggyjggMasegg$gh3ta]p!ggLhegeIMyhe t;

gggsq.yJ)1ggygg (which may be detected by the Radiation Monitor) during the time required for the signal from the Radiation Monitor to actuate the exhaust valves in full closure position. This isolation is discussed further in Section V-3,

" Secondary Containment System."

During normal plant operation, a minimum negative pressure of 0.25 inches (w.g.) is maintained by a pressure differential controller which sensen the pressure difference between outside air and the secondary containment l

atmospheres, and controls the position of the exhaust fan vortex dampers.

The i

outside air sensor has four probes, with one on each side of the building, j

Control action is initiated from the average value of the four sensor probes.

If an accident should occur, all ventilation systems of the primary containment area and secondary containment area will be isolated automatically and the SGT system will be placed in operation. Details of the SGT system are given in Section V-3,

" Secondary Containment System."

X-10-6 07/22/95

USAR 9.0 DQSE SENSITIVITY EVALUATION USING ASSUMPTIONS OF THE AEC/DRL (INCORPORATED WITH TID 14844)

NOTEi3Tho2 information presente' 51nithisTsection / wasToriginallyf intendedf to d

provide"sTeomparison1between theloriginal:1CNSl analyses,.and assumptionslin:TIDH 14844 :as Lpart ofs theloriginalf licerising fof ; the. plant. EIt-Lisithus for.0 historical

~

p'urposesionlyfandfis.;potiupdated. j 9.1 Loss-of-Coolant Accident (100 meter release height) l 1.

The reactor has operated for an extended period at 2486 MWt.

2.

100% of the noble gases in the reactor and 25% of the iodine instantaneously becomes available for leakage from the primary containment as an aerosol based on TID 14844.

3.

The primary containment volume leaks at a rate of 0.635% per day for 30 days.

4.

The escaping aerosol immediately flows through the standby gas treatment system and the stack without mixing in the secondag containment building.

5.

90% of the iodine entering the standby gas treatment is retained by charcoal filters.

6.

Meteorology For the exclusion area calculations, the concentrations are those at the plume centerline with fumigation Pasquill F Me. 2orology for the first M hour.

For the following hour and a half the conditions are slightly unstable, Pasquill C, 1 m/sec wind speed.

For the low population zone calculations, the meteorological conditions are slightly unstable, Pasquill C, 1 m/sec wind speed for the first day.

For the remaining i

29 days the conditions are 50% Pasquill D 3 m/sec wind speed. During the first j

eight hours concentrations are at the plume centerline.

During the 8-96 hour period, the plume stays within a 22M* sector. The plume stays within the 22M' sector for 1/3 of the remaining 26 days.

7.

There is a ground reflection factor of 2 for the plume and there ic no ground deposition or rain wash out of the plume.

8.

The breathing rate is 345 cc/sec for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 175 cc/sec for the next 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, and 232 cc/sec thereafter.

9.2 Refuelina Accident (100 meter release height)

(-

1.

Assumptions

'1',

'4',

'5',

'7',

and

'8' of the loss-of-coolant l

accident.

2.

Each damaged fuel rod contains 50% more activity than the average fuel rod in the core.

3, 20% of the noble gases and 10% of the iodine contained within the damaged rods are released within two hours.

4.

90% of the iodine released from the rods is retained by the refueling pool water.

XIV-9-1 07/22/96

ATTACHMENT 3 to NLS980122 COOPER NUCLEAR STATION NRC DOCKET NO. 50-298, LICENSE DPR-46 l

USAR CHANGES PREVIOUSLY MADE FOR INFORMATION ONLY AFFECTED PAGES:

1-8-2 V-3-2 through V-3-5 XIV-6-44 through XIV-6-49 For your convenience, the areas which were previously changed have been " red-lined."

l l

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l

USAR TABLE I-8-1

SUMMARY

OF MAXIMUM OFF-SITE EFFECTS OF DESIGN BASIS ACCIDENTSm Maximum Off-Site Dose (rems)

Desian Basis Accident Deen-Dose Eauivalent Thyroid Control Rod Drop Accidentczi m 4. 0 x 10-'

4.4 x 10-7 Loss of Coolant Accident'83 W

2. 3 x 10-8 2.0 x 10-*

M M ? B M T T M W LT A T M E S la% 5 W $ M W hl$ 2 G G1 75:3.11 Main Steam Line Break Accident *

1. 9 x 10-'

7.0 x 10-2 (1) Maximum off-site exposures are for 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> exposures at the site boundary.

WMMEWX$REMMl?M!O3 (3) Based on an effective release height of 30 meters for Control Rod Drop Accident.

(4) Based on an effective release height of 100 meters for Loss of Coolant Accident.

M FJER$$$f1 RfliE EL M R IL MR!MRWdMf8S2LM(gpgthe?[f1Est(Qte!!g u.

immeamamenaremen _empe.nnf.ammerg_. yx m.... mra.-.r1Wi.M.. _._ase,tt.se_i..,gh_tvor.<e,100 N1.ne.t. ers rt.o_. rt:t. h...a r

e

,_ m. ~m w..

._. um.-

n-m m

a M MELM 388A.M A=5D!81% M I3392MC M 5E3 (6) Based on an effective releese height of 30 meters for the Main Steam Line Break l

Accident.

I l

i I-8-2 07/22/96

- ^ - - - -

USAR 3.3.2 Reactor Buildina f

4 The reactor building completely encloses the reactor and its pressure suppression primary containment system.

The reactor building houses the refueling and reactor servicing equipment, new and spent fuel storage facilities and other reactor auxiliary and service equipment.

Also housed within the reactor building are the Core Standby Cooling Systems, Reactor Water Clean Up demineralized system, Standby Liquid Control system, Control Rod Drive system, instrumentation for Reactor Protection System and electrical equipment components.

The structural design features of the reactor building are described in Section XII.

Discussions of the reactor buildin 's Class I design are included in Section XII and Appendix C.

The reactor b 1 ding is also designed to meet the shielding requirements discussed in Section XII 3.3.3 Reactor Buildina Isolation and Control System The reactor building isolation and control system serves to trip the reactor building supply and exhaust fans, isolate the normal ventilation system and provide the starting signals for the SGT system in the event of the postulated LOCA inside the drywell or the postulated fuel handling accident in the reactor building.

Either of three Group 6 isolation signals will initiate the secondary containment system. The signals which indicate a LOCA inside the dr well are high drywell pressure or low reactor water level.

In addition,.

ra lation monitors in the operating which indicate a fuel handling acci(dent, refueling) floor ventilation exhaust plenum, initiate the operation of the system.

Secondary containment operation can also be initiated manually from the control room.

For further details, see Section VII-3, " Primary Containment and Reactor Vessel Isolation Control System."

Penetrations of the secondary containment are designed to have leakage characteristics consistent with secondary containment leakage requirements.

Electrical penetrations in the reactor building are designed to withstand normal environmental conditions and to retain their integrity durin the postulated fuel handling accident and LOCA inside the drywell.

Two seale doors on the equipment and personnel access locks are interlocked so that only one set of doors can be open at one time; these assure that building access cannot interfere with maintaining the secondary containment integrity.

All normally open drains are provided with water seals to maintain containment integrity.

If the Reactor Building was isolated due to high drywell pressure or low reactor water level, and the isolation signal has been reset, an additional manual action is required to reset the Group 6 isolation.

Similary, if the

~ Reactor Building was isolated due to a radiation signal, and the isolation signal can and has been reset, an additional manual action is required to reset the Group 6 isolation.

3.3.4 Standbv Gas Treatment Svstam The Standby Gas Treatment (SGT) system consists of two identical, parallel air-filtration assemblies completely enclosed within a Class I structure.

Each of the filtration assemblies is full capacity.

Each of the two SGT subsystems consists of an air tigiht sheet metal housing containing the following equipment in air flow series:"

Moisture separators - (2)

Roughing Filter - (2)

Air Heater - (1) Electric Duct Type 7.8 kW HEPA Filter - (2)

Charcoal Filters a) 6 cells per subsystem b)

Each cell 445 CFM, 2" Bed depth, 60 lb charcoal c)

Total weight 360 lbs HEPA Filter - (2)

A centrifugal exhaust fan with V-Belt drive is provided for each SGT subsystem.

USAR l

The total time required to switch from a normal containment ventilation p,ystem to the SGT s,y@ stem upon detection of a high radiation signal is lessEthsn s

rvedpisUtF90MMo $i3Keefon3hMmsxiiium3alvelclosureCt'im~eicomparedato;SG't M y typ3imet f SGT Fan Startup Time to Required Speed 15 seconds

@Z$N61aM6fCYd[v6h3p6f5(i'QnDimeEMEM.ju2 $90!s66nds The time required to bring the containment to the design negative pressure should be approximately zero assuming no accident of magnitude occurs to develop a positive pressure. Up to the time of detection of a high radiation j

i si nal the building was maintained at a constant negative pressure.

In the time i

of.lessllhan[9A3hid454d a building of this large volume would not realize a i

change An design pressure.

The activation of the second train (unit) and closing of appropriate dampers in the first train is automatic upon failure of the first train. Failure of the second train to go into service annunciated in the main control room.

1 Alarms in the main control room occur upon (a) high relative humidity I

ahead of the carbon bed, (b) low flow rate interlocked with fan "run" signal of effluent from the system, (c) high radiation in the off-gas stack (elevated release point) resulting from carbon bed or filter failure, or overload.

During primary containment isolation, each train has the capability of maintaining the secondary containment areas at an average negative pressure of 0.25 inches (w.g.) under neutral wind conditions (greater than 2 mph but less than 5 mph).

Each SGT subsystem is periodically tested as specified in the Technical Specifications.

Drywell and torus purge exhaust can also be directed to the SGT system for processing before release up the ERP (see Section II, 2.0).

The High Pressure Coolant Injection system (HPCIS) gland seal steam condenser exhauster discharge is also routed to the SGT system.

The reactor building heating and ventilation system is discussed in Section X-10.

If SGT was started due to high drywell pressure or low reactor water level, and the isolation signal has been reset, an additional manual action is required to reset the Group 6 isolation. Similary, if SGT was started due to a radiation signal, and the isolation signal can and has been reset, an additional manual action is required to reset the Group 6 isolation.

3.3.5 Elevated Release Point The location of the ERP is shown on Figure I-1-1.

The structural design of the ERP, which is 325 ft. high, is discussed in Section XII.

3.4 Safetv Evaluation The secondary containment system provides the principal mechanisms for the mitigation of the consequences of an accident in the reactor building.

The primary and secondary containment act together to provide the principal mechanisms for the mitigation of the consequences of an accident in the drywell.

If the leakage rate of the building is low, and the leakage air is filtered and discharged to the ERP, (utilizing the SGT system), the offsite radiation doses that result from postulated accidents are reduced significantly.

The reactor i

building is a Class ] structure designed in accordance with all applicable codes.

Design of the reactor building for a maximum inleakage rate of 100% per day at a building subatmcapheric pressure of 0.25 inches of water at neutral wind conditions rssulta in a low exfiltration rate even during high wind conditions.

In the event of a pipe break inside the primary containment or a fuel hsndling accident, reactor building isolation will be effected and the SGT system will be initiated. Both SGT system exhaust fans will start automatically and run to reduce the reactor building pressure to a level of -0.25 inches of water. At this point, one fan is stopped and the remaining fan exhausts to the ERP an amount equivalent to 100% per day of the reactor building free volume. With the reactor building isolated, each fan in the SGT system has the capability to hold the building at a subatmospheric pressure of -0.25 inches of water when drawing air from the building at a flowrate equivalent to 100% of the reactor building free volume per day.

Automatic exhaust fan inlet vane controls on each fan are provided to maintain the required flow rate.

V-3-4 07/22/95 w

l USAR i

i i

1 The reactor building isolation and control system performs the required isolation actions of the secondary containment system following receipt of the appropriate isolation signals.

Following initiation, the air operated reactor building isolation dampers close within approximately 12Tsili6cndpiEaddithe

~ Mr8TPpWatedH@MgdlMTdihMW64tichidainpessT'c1Tshdlw;ithisapptparilmate.ly

?

j i

Istppn#sA13The reactor building isolation control system also automatically

{

trips the reactor building and MG set supply and exhaust fans and starts the SGT system.

Section XIV-6.4 analyzes a fuel handling accident with release to the reactor buildinp. l'hisfasilysisTc"osfiriiil"tWiit@usiusiFgTan"iisfiltisied ~itelsisis

. CliNiiihtbr i " 3eshaustefanMflownfork appipokimately190 Misecondst

$wg1Pl

. ationivigtihssmoth piratpdLdangierdfollowe'd by 4 T;is.ystema releasess @renwellibelow2theisitelboundary aceleaseg a

_esipermi'tted3yMQCFR1n00j The SGT system filters exhaust air from the reactor building and discharges the processed air to the ERP.

The system filters particulate and iodines from the air stream in order to reduce the level of airborne contamination released to the environs via the ERP.

When the system is exhausting from the reactor building, the building is held at a minimum i

subatmospheric pressure of -0.25 inches of water.

The ERP provides a release for airborne activity during the postulated station loss of coolant and refueling accidents.

Release of activity i

to the environs from the secondary containment system is analyzed in detail in l

Section XIV, Station Safety Analysis.

It is concluded that the safety design bases are met.

l 3.5 Insnection and Testinc i

The secondary containment leakage rate can be determined in the i

l following manner. The reactor building is isolated and the SGT system is started f

with one treatment train and its associated exhaust fan.

The exhaust flow rate j

is controlled by the fan inlet vane control position as determined by flow rate measurements in the SGT system exhaust duct.

The fan inlet vane positioner is used to control the exhaust flow rate at a volume of 100% per day of the reactor i

building free volume.

If the subatmospheric pressure as measured within the reactor building is equal to or exceeds 0.25 inches of water (with neutral wind conditior.s at the site) the building safety design basis leaktightness with respect to inleakage is verified.

Tests of the ability of the various isolation initiation signals to automatically render the reactor building isolated, to trip the supply and exhaust fans and to start the SGT system can be conducted by simulating the isolation signals.

I Provisions are made for periodic tests of each SGT filter unit.

l These tests include, at a minimum, determinations of differential pressure across each filter and of filter efficiency. Connections for testing, such as injection i

and sampling, are located to provide adequate mixing of the injected fluid and representative sampling and monitoring, so that test results are indicative of performance. Each HEPA filter can be tested with DOP (di-octylphthalate) smoke.

The charcoal filters can be tested for bypass with freon, j

The electric heating coil in each SGT filter train is tested and I

shown to reduce the relative humidity of an entering air stream.

i 3.6 Nuclear Safetv Operational Recoirements NOTE: Limiting Conditions for Operation and Surveillance testing i

requirements stated and listed in this subsection are based on analyses performed at the time of original license application. For current information refer to the Technical Specifications.

General Table V-3-1 represents the nuclear safety operational requirements for the secondary containment system for each BWR operating state.

The entries in Table V-3-1 represent an extension of the plant-wide BWR systems analysis of Appendix G.

The following referenced portions of the USAR provide important information justifying the entries in Table V-3-1:

V-3-5 07/22/96 l

I I

USAR i

j TABLE XIV-4-2 STATION SAFETY ANALYSIS RESULTS OF DESIGN BASIS ACCIDENTS Maxinun 24-Hour Off-Site Dose (rems)****

I Percent of Core

)

Reaching Cladding Temperature of Peak Nuclear System Deep-Dose Design Basis Accident 2200*F Pressure Equivalent Thyroid Rod Drop Accident 0

<1375 psis 4.0x10 4.4x10' 4

Less of Coolant Accident 0

Not applicable

• 2.3x10'8 2.0x10 j

d 4

R2 fueling Accident 0

Not applicable **

4Mx10-gy Main Steam Line Break Accident 0

Not applicable

• 1.9x10'3 7.0x10'8 Loading Error Accident 0

Not applicable 0***

0***

One Recirculation Punp Seizure 0'

Not applicable

• 0***

0***

Accident

• This accident results in a depressurization.

• • This accident occurs with the reactor vessel head off.

• • • No fuel failures are predicted.

• • • • Based upon analysis of 7x7 fuel bundles, j

i l

l xlv 4-4 07/22/96

USAR 6.4.8.2 Fission Product Release to Secondarv Containment The following assumptions and initial conditions are used to calculate the fission product release to the secondary containment.

a.

The fission product activity released to the secondary containment will be in proportion to the removal efficiency of the water in the refueling pool.

Since water has a poor affinity for the noble gases they are assumed to be instantaneously released from the pool to the secondary containment.

b.

As noted in Section XIV-6.3.8.1, the removal efficiency of the water for halogens can be defined in terms of the partition factor, for which values between 102 and 10' have been experimentally determined to be applicable for the conditions under investigation.

A partition factor of 102 for the halogens has been conservatively assumed for this accident.

Thus the computed inhalation doses will be overestimated by a factor of 10 to 10.

5 c.

The conservative assumption is also made that instantaneous equilibrium is attained between the refueling pool and secondary containment.

In reality, if a true equilibrium is maintained, the effects of plateout or fallout would be compensated for by the evolutions of activity from the refueling pool.

d.

The effects of plateout and fallout are neglected.

Fission product plateout and/or fallout will occur in the secondary containment; however, for the assumption that a true equilibrium is maintained, the effects of plateout or fallout would be compensated for by the evolutions of activity from the refueling pool.

e.

The refueling cavity liquid volume is 3. 0 x loi f t' and the ef fective air volume in the secondary containment is 7.95 x 10' f t'.

f.

The Standby Gas Treatment System removes 1 secondary containment air volume per day.

Based upon these assumptions, the airborne activity is as shown in Table XIV-6-8.

6.4.8.3 Fission Product Release to Environs The following assumptions and initial conditions are used to calculate the fission product release to the environs.

j a.

High radiation levels in the reactor building exhaust plenum will isolate the normal reactor building and MG set ventilation systems, and actuate the standby gas treatment system.

It ~ is cassumedi that cit: ' takes approximately.901 seconds 9toEisolatenthe; reactor; building ; During'the; period, full:exhaus_t7 flow:from;the(operating; reactor building ventilationLexhaust; fan;is ct.nservatively; assumed,gresulting f n ; an unf1.ltered release f rom. the ; reactor l

i building; roof; 1

I i

XIV-6-44 07/22/96 i

l a_-________.

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XIV-6-45 07/22/96 u________________-_______.__.

USAR b.

The relative humidity in the secondary containment is 70 percent. Since the refueling accident does not result in the release of any liquid or vapor to the secondary containment, the normal environmental condition existing prior to the accident will also exist after the accident, except for the addition of the released fission products.

The relative humidity in the secondary containment will therefore be considerably below any levels which may be detrimental to the filter media in the Standby Gas Treatment System. However, as mentioned previously, the charcoal beds and absolute filter media, as well as the air flowing through the filter system, are heated 10*F above the mixture entering the system, reducing the relative humidity to 70 percent or less.

The filter efficiency is assumed to be 99 percent for iodines c.

and 0 percent for the noble gases (see Subsection XIV-6.3.6c).

d.

There is 1 secondary containment air change per day through the Standby Gas Treatment System. The activity in the secondary containment is shown on Table XIV-6-8.

Based upon these conditions, the fission product activity release rate to the environs is as shown in Table XIV-6-9.

I 6.4.8.4 Radiological Effects The radiological doses to the general population have been evaluated for six meteorological conditions ranging from very stable to unstable meteorology occurring with 1 and 5 meters per second winds.

Two dose periods have been evaluated, a 2-hour dose period and a 24-hour dose period, commonly referred to as the total dose.

Two hour doses were calculated using two hour I

atmospheric dispersion values.

Twenty-four hour doses were conservatively calculated, also using two hour atmospheric dispersion values.

It should be emphasized that the radiological doses presented in Tables XIV-6-10a and XIV-6-10b are based upon the assumption that the stated meteorological conditions exist for the duration under consideration and that the wind blows in one direction during the entire release period.

I Tables XIV-6-10a and XIV-6-10b show the calculated off-site doses beyond the nearest site boundary (approximately 725 m from the release point).

The values in these tables assume a flat site.QThe]4-houriwholeibodyland thyroidlinhalationidosesiareJ4J1*x310 sremyandygrespectivelyf JThese" doses /

4 asyelliasychose;shown;ongTablesiXIVn6310aland;XIV-6t1_Ob;Eare"below the; guideline q

wholey,bodypandlylif etimeythyroid %dosesgofy 252rempandy300grem q(10 CFR100 ) 1; respectively Section 15.3 of the CNS SER calculated the two-hour raalaticn l

exposure at the nearest site boundary for this accident to be less than 1 rem to the thyroid and less than 1 rem whole body.

In determining these doses, it was assumed that 10 percent of the noble gases and 10 percent of the halogens were released from the ill damaged rods. Only 1 percent of the halogens released from the fuel are released from the pool water. Halogens released from the pool water are 25 percent organic and 75 percent elemental. The airborne fission products are discharged via the standby gas treatment system (iodine removal ef ficiency of 90 percent for elemental forms and 70 percent for organic forms) from the 100 meter stack over a two-hour period.

It is concluded that this accident will not result in any radiological doses which endanger the health and safety of the public.

XIV-6-46 07/22/96 i

USAR i

TABLE XIV-6-9 i

REFUELING ACCIDENT (7x7 FUEL BUNDLES)

INTEGRATED FISSION PRODUCT RELEASE TO THE ENVIRONS Time After Noble Gases Iodines Accident (Curies)

(Curies) l i

s ues 64 '

I PtTxEM E ls a g gyygggs EFM ER5MIM' EIM m M' i.

1 i

1 I

i i

t XIV-6-47 07/22/96

USAR TABLE XIV-6-10a REFUELING ACCIDENT (7x7 BUNDLES)

RADIOLOGICAL EFFECTS 2-HOUR DOSE DISTANCE METEOROLOGICAL CONDITIONS (METERS)

VS-1 MS-1 N-1 N-5 U-1 U-5

)

PASSING CLOUD DEEP-DOSE EQUIVALENT (REM)

(

725*

gg3 plag}

$233 1343g 3',;3g33 6,5;}*4

)

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g'g1EjB, ggSJEg

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s 3218 gg BR?gjB,

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gggE gj]IO 3J6];EO ggIES

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[gg(2 gg(E-;1, Q"B,g 2EEg1;3f;2 3p]Q 16090 [75jgg 213AEg 2135Eg g23TER QlE;3, Sj51EM '* Site Boundary Mandford Meteorolocrv Wind Soeed (M/S) i VS-1 Very Stable 1 l MS-1 Moderately Stable 1 I N-1 Neutral 1 N-5 Neutral 5 U-1 Unstable 1 i U-5 Unstable 5 l l i XIV-6-48 07/22/96 f A

USAR TABLE XIV-6-10b REFUELING ACCIDENT (7x7 FUEL BUNDLES) RADIOLOGICAL EFFECTS 24-HOUR DOSE DISTANCE METEOROLOGICAL CONDITIONS M TERS) VS-1 MS-1 N-1 N-5 U-1 U-5 PASSING CLOUD DEEP-DOSE EQUIVALENT (REM) 725* gggg - g2[gg giQg gI33Ef! M;*EZEM gyp &RQ 1050 reg g[OJE' 2 2:4gEQ g!LES gZ4jJg g10JEM 9 1609 ggg 2fp!B 2 E7gg 3153BQ gjlEg 1103g j 3218 Etgrjg 1938 gJgg

M63fS,

}ggf3 g]533 8045 E g3 ((7J,EM 21gj3 Q3;4, 316]Egd, D6gEf! 16090 338JJM gja EM gplEM $3JEf4 170?JEj,3 2 M Eg5 j l LIFETIME THYROID DOSE (REM) 725* M5tE?6 113 P3.~0 2_17.J50 576 -,F. E. xv1 8755 n-.5s.1 R.i2,u a*1 FE k u r .. u ~ 1050 EggggpEgggLDat:41NotMgiMle, ggigg i 1609 ggM &# 5;_30 g* g g"2MM Rg[E-3 SjfjaEt2 J 4 3218 gggg ggg g y/y} ggEg gigf,E-3 }j.QEp 8045 [TS *SW3, g_7,.1_?lE.f.1 ET.9T1B*w 2._74"E.m_2 1737Ee2 314.~5 E.. e3.. u2 - -~ ~ r 16090 2776 ? B31.. 2Y4REa m .a% & -.a_m.1 2.. Tie rE..m., 7 T.5.~7E.a,- 37*f 3E-3 2 d3 o.7 0. r E...- 3 1 _m*. Me._tero._lDEUha, n.~d.rWi.nd, CSpe. e.dt D.a.. tax. i_,s?id._e._nti. ~ca..l3,t_o. ?.,thatFi._nm R.eferenc.m.W Stable?37 m - ~ e L ~- m m g.m ~m m m 4 ~ CSite Boundary l Meteoroloov Wind Soeed (M/S) l VS-1 Very Stable 1 MS-1 Moderately Stable 1 N-1 Neutral 1 N-5 Neutral 5 U-1 Unstable 1 U-5 Unstable 5 XIV-6-49 07/22/96

ATTACIIMENT 4 to NLS980122 COOPER NUCLEAR STATION NRC DOCKET NO. 50-298, LICENSE DPR-46 1 SIMPLIFIED DIAGRAM OF THE AS-BUILT CONFIGURATION 1 I

+ fi si ..or se hT 25 f "~ l? .. E, ?. n'li ag 3 'i ~~ l.fr _ _!) .r., _ i s L 41 -o gg --vQ, if.'+- ~.. A u gao il s s.i ss i ...~v. g gg 5 g. 4-gg ! I) l' f i fili I yp l pI! ) s r-a A8' + I tilt iiiE- ! l 5 i, = - si3l 4-ar 3 ~ ~ II' I -i g I "M 1,, p_:t g8 - r_. p 'l fr ^ 1i8' iE-ig !s lE sf' A _ill !I 8 olf 4 si e;f !I- "A tili insi

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l F* TAC ~IMENT 3 LIST OF NRC COMMITMENTS l Correspondence No: NLS980122 The following table identifies those actions committed to by the District in this document. Any other actions discussed in the submittal represent intended or planned actions by the District. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the NL&S Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments. COMMITTED DATE COMMITMENT OR OUTAGE The USAR will be updated to clearly reflect the existing as-built olant configuration for secondary containment isolat'va and the postulated Refueling Accident analysis results which conclude that a limited fission product Within 30 days of NRC release occurs through the normal ventilation discharge

approval, path.

Proposed USAR changes are contained in to NLS980122. Clarify appropriate sections of ITS bases to reflect the existing as-built plant configuration for secondary containment isolation and the postulated Refueling Within 30 days of NRC Accident analysis results which conclude that a limited approval. fission product release occurs through the normal ventilation discharge path. Proposed changes to ITS are contained in Attachment 2 to NLS980122. Restrictions on movement of fuel and heavy loads over Until NRC approval of irradiated fuel will remain in place until approval of proposed license the proposed license amendment (NLS980122). amendment. l l l PROCEDURE NUMBER 0.42 l REVISION NUMBER 6 l PAGE 9 OF 13 l _ _ _ _ _ _ _ _ _ _ _ _ _}}