ML20117E782

Application for Amend to License NPF-58,revising MSL Leakage Requirements & Eliminating MSIV Leakage Control Sys. Attachment 7 Provides Proprietary Dose Calculations. Attachment 7 Withheld

ML20117E782
Person / Time
Site:	Perry
Issue date:	08/27/1996
From:	Shelton D CENTERIOR ENERGY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML19311C241	List: ML20117E782 ML20117E789 ML20117E810 PY-CEI-NRR-2076, Rev 2 to Nonproprietary Version of Controlled Copy of Dose Calculation Data Base for Application of Revised DBA Source Term to CEI Pnpp
References
RTR-NUREG-1465 PY-CEI-NRR-2076, NUDOCS 9609030089
Download: ML20117E782 (25)

Text

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i w ssesacy PERRY NUCLEAR POWER PLANT Mail Address: Donald C. Shelton 10 CENTER ROAD SENIOR VICE PRESIDENT PERRY. OHIO 44081 PfRR .O 10 44081 NUCLEAR (216) 259-3737 j I

l August 27,1996  !

PY-CEI/NRR-2076L l

United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Perry Nuclear Power Plant  :

i Docket No. 50-440 License Amendr '.equest: Revision of Main Steam Line Leakage Requirements and i Elimination of tr.; ain Steam Isolation Valve Leakage Control System I Gentlemen: 1 Amendment of the Facility Operating License (NPF-58) for the Perry Nuclear Power Plant (PNPP) is requested. The proposed changes affect Technical Specification requirements for Main Steam Line leakage and the Main Steam Isolation Valve Leakage Control System (MSIV-LCS). PNPP is a pilot plant in the collaborative efforts of the Nuclear Regulatory Commission (NRC), the Nuclear Energy Institute (NEI), and the Electric Power Research Institute (EPRI) for implementation of the NRC research documented in NUIEG-1465,

" Accident Source Terms for Light-Water Nuclear Power Plants". NUREG-1465 is based on over 20 years of more realistic accident source term studies. The proposed changes are based on reanalysis of the design basis Loss of Coolant Accident using the revised accident source term from NUREG-1465 and the NEI document entitled " Generic Framework for Application of Revised Accident Source Term to Operating Plants". The proposed amendment was the subject of a meeting with the NRC staff on May 30,1996.

This pro; ed change has several safety benefits. Due to the mechanistic characterization of the fissior. mduct source term in NUREG-1465, more appropriate mitigative methods may be employed post-accident. In addition, the proposed increase in allowable Main Steam Line leakage will reduce the frequency of Main Steam isolation Valve (MSIV) maintenance, thereby resulting in radiation dose reductions to maintenance workers. The proposed change also has a significant cost benefit. This change is expected to result in cost reductions well in excess of $100,000, which qualifies it as a cost beneficial licensing action.

Implementation of this license amendment prior to restart from the next refueling outage (refuel outage 6) will fulfill a commitment made to the NRC in a letter dated April 25,1995 30 O () g "9 (PY-CEl/NRR-1934L). The commitment is to resolve issues which result in dec inboard MSIV-LCS subsystem inoperable during plant startups, prior to restart from refuel l outage 6. Currently, an exception to Limiting Condition for Operation (LCO) 3.0.4 is \p operotmg compon.es Clevelona Electnc Hlummotmg i '[

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incorporated into the Technical Specification for the MSIV-LCS, because until plant power levels are increased above 50 percent, the inboard MSIV-LCS subsystem is considered to be inoperable (further details on this issue are in the April 25,1995 letter and in License Amendments 63 and 71). The existing 3.0.4 exception was issued by Amendment 71 to the Operating License, and will expire upon completion of the sixth cycle of operation.

Elimination of the MSIV-LCS Technical Specification requirements beginning with refuel outage 6 (as proposed by this licensing submittal) will resolve the issues which result in declaring the inboard MSIV-LCS inoperable during operations below 50 percent rated thermal power and, consequently, will eliminate the necessity for the 3.0.4 exception.

Attachment 1 provides a Summary, a Description of Proposed Changes, a Safety Analysis, and an Environmental Consideration. Attachment 2 provides the Significant Hazards Consideration. Attachment 3 provides a copy of the marked-up Technical Specification pages. Attachment 4 provides marked-up Table of Contents and Bases pages, for information, since these documents are not part of the Technical Specifications. Attachment 5 provides information on a study which served as the basis for developing site specific atmospheric dispersion values for Control Room dose calculations. Attachment 6 provides the nonproprietary dose calculations. Attachment 7 provides the proprietary dose calculations.

To support implementation in refuel outage 6, we request that this amendment be issued by May 1997.

If you have questions or require additional information, please contact Mr. James D. Kloosterman, Manager - Regulatory Affairs at (216) 280-5833.

Very truly your holdf '

fo onald C. Shelton DAY:dy Attachments cc: NRC Region 111 Administrator NRC Resident inspector NRC Project Manager State of Ohio

I, Richard D. Brandt, being duly sworn state that (1) I am General Manager, Perry Nuclear Power Plant Department of the Cleveland Electric Illuminating Company, (2) I am duly authorized to execute and file this certification on behalf of The Cleveland Electric Illuminating Company and Toledo Edison Company, and as the duly authorized agent for Duquesne Light Company, Ohio Edison Company, and Pennsylvania Power Company, and (3) the statements set forth herein are true and correct to the best of my knowledge, information and belief.

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( Richp W D. Brandt V

S r to and subscribed before me, the [ 7d' day of FA-(<2I~ ,

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r l s ATTACHMENT 7 CONTAINS PROPRIETARY INFORMATION.

PER THE ATTACHED POLESTAR PROPRIETARY LETTER,IT SHOULD BE HANDLED IN ACCORDANCE WITH 10CFR2.790

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1 J%uumes r APPUEo Tecwcroav e July 18,1996 D^v' E . W. LEAVE R Mr. Emin Ortalan Cleveland Electric Illuminating Company Mail Zone A-170 10 Center Road Perry,OH 44801

Dear Mr. Ortalan:

Attached is a Polestar Affidavit dated July 18,1996 which sets forth exemptions in accordance with 10 CFR 2.790(a)(4) which justify the non-disclosure of portions of Polestar reports and calculations developed in support of the Perry Nuclear Plant application of the revised accident source term for DBA. This Affidavit applies to the recent revision to the source term calculation package and is intended to replace the Affidavit dated April 25,1996.

On the basis of this non-disclosure justification, it is requested that CEI not publish or otherwise disclose the proprietary version of the Polestar reports and calculations, other than for the purpose stated in the Polestar - EPRI contract on the Perry revised source term work, without written consent of Polestar.

A non-proprietary version of the latest revision of the calculation package, for l which there is no restriction on disclosure, has been sent to you under separate cover.

l Please contact me with any questions or need for further information.

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Very truly yours, l

ONE FiRsr STREET

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• Los altos C ALIF ORNIA

• 94022 . USA . TEL 415-948 8242 . FAv 415-948-8244

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Polestar Applied Technology,Inc.

DAVio E. W. LEAVER I

AFFIDAVIT I I, David E.W. Leaver, being duly sworn, depose and state as follows:  !

(1) I am a Principal and an Officer of Polestar Applied Technology, Inc.

(" Polestar") and am responsible for the function of reviewing the ,

information described in paragraph (2) which is sought to be withheld, and l have been authorized to apply for its withholding.

l (2) The information sought to be withheld is contained in portions of the following Polestar reports and calculations prepared for CEI in support of the Perry Nuclear Plant application of the revised design basis accident source term:

PSAT 04212H.02, Drywell Sweep-Out Rate and Related Thermal-Hydraulic Conditions Inside Containment PSAT 04202H.04, Aerosol Decay Rates (Lambda) in Drywell PSAT 04202H.05, Aerosol Decay Rates (Lambdas) in Containment with Spray PSAT 04212H.06, Mixing Between the Sprayed and Unsprayed Portions of the Perry Containment PSAT 04202H.07, Main Steam Line Heat Transfer Analysis PSAT 04202H.08, Steamline: Particulate Decontamination Calculation PSAT 04202H.11, Perry Containment Water Pool pH PSAT 04202H.12, Calculation of Fraction of Containment Aerosol Deposited in Water -

(3) In making this application for withholding of proprietary information of which it is the owner, Polestar relies upon the exemption from disclosure set forth in the NRC regulations 10 CFR 9.17(a)(4),2.790(a)(4), and 2.790(d)(1) for

" trade secrets and commercial or financial information obtained from a person and privileged or confidential"(Exemption 2.790(a)(4)). The material for which exemption from disclosure is here sought is all " confidential commercial information".

1 ONE Fisst S1Hr ET . Sun e 4 . Los altos . C AL IFORNIA . 94022 . US A . TEL 415-948-8242 . fax 415 948-8244 i

9 (4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process or method, including supporting data and analyses, where prevention of its use by Polestar's competitors without license from Polestar constitutes a competitive economic advantage over other companies.

b. Information which,if used by a competitor, would significantly reduce his expenditure of resources or improve his competitive position in the analysis, design, assurance of quality, or licensing of a similar product;

c. Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of Polestar, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future Polestar customer-funded development plans and progrcms, of potential commercial value to Polestar;

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in both paragraphs (4)a and (4)b, ab we.

(5) The information sought to be withheld is being submitted to CEI (and, we trust, to NRC) in confidence. The information is of a sort customarily held in confidence by Polestar, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Polestar, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Distribution of such documents within Polestar is limited to those with a need to know.

~ (7) The approval of external release of such a document typically requires review by the project manager, and the Polestar Principal closest to the work, for 2

m i l l technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside Polestar are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and j licensees, and others with a legitimate need for the information, and then l only in accordance with appropriate regulatory provisions or proprietary l agreements.

l (8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed information on models in current software which were developed by Polestar and are justified from numerous benchmark exercises and applications, detailed results of these models and analytical methods, and computer codes which Tylestar has developed, documented, and is maintaining under the Polestar 10 CFR 50, Appendix B Quality Assurance Program. The model results are from applications of the revised DBA source term to the Perry Nuclear Plant in support of CEI.

The development and documentation of the methods, models, and associated -

Polestar computer codes used in these analyses was achieved at a significant cost to Polestar, on the order of $50,000, which is a significant fraction of internal research and development resources available to a company the size of Polestar.

The development of the models and methods, along with the interpretation and application of the results,is derived from the extensive experience database that constitutes a major Polestar asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to Polestar's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of Polestar's comprehensive technology base on application of the revised source term to operating plants and advanced light water reactors, and its l

commercial value extends beyond the original development cost. The value ,

of the technology base goes beyond the extensive physical database and j analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with methods which have been developed and are being maintained in accordance with 10 CFR 50, Appendix B requirements.

i l The research, development, engineering, analytical and review costs comprise a substantial investment of time and money by Polestar.

l The precise value of the expertise to devise an evaluation process and apply '

the correct analytical methodology is difficult to quantify, but it clearly is substantial.

3

l Polestar's competitive advantage will be lost if its competitors are able to use l the results of the Polestar experience to normalize or verify their own process i or if they are able to claim an equivalent understanding by demonstrating l that they can arrive at the same or similar conclusions.

The value of this information to Polestar would be lost if the information were disclosed to the public. Making such information available to i

competitors without their having been required to imdertake a similar expenditure of resources would unfairly provide competitors with a windfall, l and deprive Polestar of the opportunity to exercise its competitive advantage i

to seek an adequate return on its relatively large investment in developing these very valuable analytical tools.

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STATE OF CALIFORNIA )

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COUNTY OF SANTA CLARA )

David E.W. Leaver, is being duly sworn, deposes and says:

That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge,information, and belief.

Executed at Los Altos, California, this 1996.

/ day of M>  ;

i U EN David E.W. Leaver Polestar Applied Technology, Inc.

i Subscribed and sworn before me this /f day of I 1996.

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Attachment 1 PY-CEI/NRR-2076L l Page1 of 9

SUMMARY

Removal of the Technical Specification requirements for the Main Steam Isolation Valve Leakage Control System (MSIV-LCS), and increasing the allowable leak rate specified for the Main Steam lines is the result of applying the revised accident source tenn (as documented in NUREG-1465 and the Nuclear Energy Institute (NEI) document " Generic Framework for Application of Revised Accident Source Term to Operating Plants") to the design basis Loss of

, Coolant Accident (LOCA) off-site and Control Room dose analysis for the Perry Nuclear Power  ;

Plant (PNPP). 1 The proposed amendment was the subject of a meeting with the NRC staff on May 30,1996.

MSIV-LCS l

Technical Specification 3.6.1.9 identifies the operability requirements for the MSIV-LCS. I l

In 1984, a program was initiated by the NRC to make regulatory requirements more efficient by l eliminating those with marginal impact on safety. An industry survey resulted in a list of 45 candidates for potential regulation modification. The survey results and analyses of the selected ,

candidates were published in NUREG/CR-4330 " Identification of Regulatory Requirements that may have Marginal Importance to Risk", Volumes 1,2, and 3. One of the candidates was to  ;

eliminate the requirement for the MSIV-LCS in Boiling Water Reactors (BWRs) per Regulatory Guide 1.96, " Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water Reactor Plants". This submittal is a result of continuing efforts to achieve the goal of eliminating that system.

Removing the Technical Specification requirements for the MSIV-LCS is based on reanalysis of off-site and Control Room doses, where the MSIV-LCS is not credited in the calculation. As noted above, the reanalysis utilizes the revised design basis accident (DBA) source terms. The l limiting reanalysis case assumes that main steam line leakage is attenuated in the main steam line from the reactor vessel out to the outboard MSIV. This is the limiting scenario since the worst case single failure, and hence the most limiting analysis case, involves a failure to close the valve downstream of the outboard MSIV in each main steam line, i.e., the Main Steam Stop Valves (INIIF0020A,B,C and D). Although this most limiting analysis case assumed a failure to close the Main Steam Stop Valves, retention of OPERABILITY requirements on these valves is appropriate to ensure the single failure analysis remains valid.

Not crediting the MSIV-LCS in the design basis accident analysis is consistent with the approach taken by several BWR licensees, which have applied for NRC approval of this change using an approach developed by the Boiling Water Reactor Owners Group (BWROG). The BWROG methodology involves seismically qualifying the main steam lines out to and including the non-safety related, non-seismic drain lines and main condenser, and then using that volume to attenuate leakage past the MSIVs. At PNPP, the existence of safety related,

Attachment 1 PY-CEI/NRR-2076L Page 2 of 9 seismically qualified piping leading to the safety related, seismic, Class 1E powered Main Steam Stop Valves, together with the characteristics of the revised accident source term (i.e., predominantly aerosol which is largely retained in the drywell, containment and main steam lines) provides the option of taking credit only for the volume within the main steam lines for leakage attenuation.

Although the requirements for the MSIV-LCS are being removed (since credit is no longer taken for the system as part of the design basis accident analysis), OPERABILITY requirements on the Main Steam Stop Valves are being retained since the valves meet Criterion 3 of 10CFR50.36 (c)(2)(ii). Specifically, a technical specification limiting condition for operation must be established for each item that is "A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis ,

accident or transient that either assumes the failure of or presents a challenge to the integrity l of a fission product barrier." The only necessary surveillance requirement is to ensure the j valves will stroke closed on a manual demand by the operators. Leak test requirements are not necessary to ensure the assumptions of the dose calculation methodology are met for the

)

main steam lines, since leakage flow characteristics used in the analyses are affected only by the turbulence caused by an open ended pipe (i.e., the Main Steam Stop Valves fail to close).

Implementation of this license amendment prior to restart from the next refueling outage (refuel outage 6) will fulfill a commitment made to the NRC in a letter dated April 25,1995 (PY-CEI/NRR-1934L). The commitment is to resolve issues which result in declaring the inboard MSIV-LCS subsystem inoperable during plant startups, prior to restart from refuel outage 6. Currently, an exception to Limiting Condition for Operation (LCO) 3.0.4 is incorporated into the Technical Specification for the MSIV-LCS, because until plant power levels are increased above 50 percent, the inboard MSIV-LCS subsystem is considered to be inoperable (further details on this issue are in the April 25,1995 letter and in License Amendments 63 and 71). The existing 3.0.4 exception was issued by Amendment 71 to the Operating License, and will expire upon completion of the sixth cycle of operation.

Elimination of the MSIV-LCS Technical Specification requirements beginning with RFO 6 (as proposed by this licensing submittal) will resolve the issues which result in declaring the inboard MSIV-LCS inoperable during operations below 50 percent rated thermal power and, consequently, will eliminate the necessity for the 3.0.4 exception.

The physical isolation of the MSIV-LCS from the Main Steam system will eliminate leakage pathways. This modification will be performed as part of the PNPP design change process.

MAIN STEAM LINE LEAKAGE RATE This amendment request proposes to limit the leakage through each main steam line to less than or equal to 100 scfh, as long as the combined leakage rate through the four main steam lines is less than or equal to 250 scfh. Technical Specification 3.6.1.3 currently requires that the main steam line leakage rates shall be limited to less than or equal to 25 standard cubic feet per hour (scfh) through each main steam line when tested at the calculated peak containment pressure

i Attachment 1 PY-CEI/NRR-2076L Page 3 of 9 (Pa). [ Note: Until the end of Operating Cycle 6, an exception to the 25 scfh limit is provided for l one of the main steam lines. Specifically, the leakage rate through one of the main steam lines is limited to less than or equal to 35 scfh (versus 25 scfh), provided the total leakage through the ,

four lines remains less than or equal to 100 scfh]. l The purpose for limiting the main steam line leakage rate is to ensure isolation of the reactor coolant system in the event of a design basis LOCA. Industry operating experience has shown that these valves invariably exhibit some level of minor leakage. The current Technical Specification allowable leakage rate is extremely small considering the physical size and operating characteristics of the MSIVs (i.e., large size and fast acting). Based on an in-depth evaluation of MSIV leakage (refer to NEDC-31858P, "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems", Revision 2, and the summary in NUREG-1169 " Technical Findings Related to Generic Issue C-8; Boiling Water Reactor Main Steam Isolation Valve and Leakage Treatment Methods"), the BWROG has concluded that leakage rates of up to 500 scfh are not indicative of substantial mechanical defects in the valves which would challenge the valves capability to fulfill their safety function ofisolating the steam lines. Therefore, as demonstrated in the design basis LOCA radiological reanalysis, the proposed increased allowable main steam line leakage rate (each line less than or equal to 100 scfh, and total leakage less than or equal to 250 scfh, when tested at Pa) will not affect each MSIV's isolation function capability. Therefore, the overall level of plant safety will be maintained, while reductions can be achieved in: radiation dose to maintenance workers; the maintenance workload during plant outages; and the potential for outage extensions.

Per agreements reached with the NRC through the BWROG program for increasing main steam line leakage rate limits, a commitment is made that if the leakage rate on a main steam line exceeds 100 scfh during Technical Specification required testing, the leakage rate for that line I

will be restored to s 25 scfh when tested at Pa (the current criterion for leakage), prior to plant restart.

DESCRIPTION OF TIIE PROPOSED TECIINICAL SPECIFICATION CIIANGE Technical Specification Surveillance Requirement (SR) 3.6.1.3.10 will be revised to limit the leakage through each main steam line to less than or equal to 100 scfh when tested at Pa, as long as the combined leakage rate through the four main steam lines is less than or equal to 250 scfh when tested at Pa. In addition, the Specification pertaining to the MSIV-LCS will be deleted. In its place, a new Specification entitled " Main Steam Stop Valves" will be created, which retains OPERABILITY requirements on the Main Steam Stop Valves to ensure that the Main Steam Stop Valves remain OPERABLE during MODES 1,2 and 3. Refer to Attachment 3 for a l marked-up copy of the affected Technical Specification pages.

i

[ Note: The Bases will be revised under the PNPP Bases Control Program to reflect the changes identified above. In addition, the Bases will be revised to provide clarifying text to the discussions on the Containment Spray (B 3.6.1.7), and Combustible Gas Mixing (B 3.6.3.3)

Systems, in order to reflect their contribution to the post-LOCA removal of airbome 4

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i Attachment 1 PY-CEI/NRR-2076L i

Page 4 of 9 radionuclides. The proposed Bases and Table of Contents mark-ups are contained in Attachment 4 "for information only".]

l This proposed change has been developed for implementation during refuel outage 6.

l SAFETY ANALYSIS i

l Removal of the Technical Specification requirements for the MSIV-LCS, and increasing the l allowable leak rate specified for the Main Steam lines is the result of applying the revised accident source term (as documented in NUREG-1465 and the NEI Generic Framework document (Reference 1)) to the design basis accident (DBA) LOCA off-site and Control Room dose analysis for the Perry Nuclear Power Plant (PNPP).

As noted in the NEI Generic Framework Document and in the PNPP LOCA reanalysis (see Reference 9 (included as part of Attachment 6)), the revised DBA source terms of NUREG-1465 are comparable in conservatism to the DBA source terms previously used at PNPP. The noble gas and iodine release fractions (which are the main determinants of the whole body and thyroid dose evaluations specified in 10CFR100) are about the same. The revised accident source term timing and chemical form, while different from the previous source terms, are nonetheless conservative compared to what is expected under actual accident conditions, (e.g.,

the 1979 accident at Three Mile Island) and provide a more physically correct representation of activity release to the containment. Knowledge of the more physically correct representation of the timing and chemical form provides the opportunity to develop the most appropriate mechanisms for mitigating radiological releases. 1 Furthermore, in terms of activity transport within and through the containment system and release to the environment, there are many other conservatisms included in the LOCA l reanalysis. The following provides a brief summary of some of the activity transport l conservatisms that exist within the dose reanalysis. These are explained more fully in Reference 9 (included as part of Attachment 6).

1. Earlier gap release start time than required for a BWR

2. Underestimated volumetric flow from the drywell during core damage / debris quench

3. Neglected suppression pool scrubbing

4. Neglected natural aerosol removal in unsprayed regions of containment

5. Underestimated containment spray effectiveness

6. Most conservative break location l

7. Most conservative distribution of total main steam line leakage

8. Most conservative main steam line valve single failure

9. Underestimated natural aerosol removal in main steam line

10. Underestimated drywell aerosol natural removal

11. Instantaneous iodine release in main steam lines l

l

Attachrnent1 PY-CEI/NRR-2076L Page5 0f 9 These are significant conservatisms, which result in calculated offsite and Control Room doses higher than those that would exist following a release of a NUREG-1465 source term from the ,

reactor core. These activity transpmt conservatisms in the LOCA reanalysis are judged to be comparable to the conservatisms utilized in the original analyses.

Although (as stated above) the revised source term is considered "more physically correct" than the previous source terms used in performing design basis LOCA radiological analyses, it should be noted that in an actual design basis LOCA, no fuel damage would occur. The plant operating limits, such as the Maximum Average Planar Linear Heat Generation Rate

) (MAPLHGR), are chosen to ensure that post-LOCA fuel cladding temperatures remain low i

enough to maintain fuel pin integrity. As discussed in USAR section 15.6.5.5.2 (the Realistic Analysis),"the only activity released to the drywell is that activity contained in the reactor coolant plus any additional activity which may be released as a consequence of reactor scram and vessel depressurization." Therefore, off-site and Control Room doses would be minimal (significantly less than the calculated values presented below, which utilize the NUREG-1465 source term and other inherent conservatisms such as the activity transport conservatisms described above). The results of the DBA LOCA off-site and Control Room dose reanalysis are  :

provided below.

DOSE RESULTS (REM)

Proposed Existing USAR USAR Regulatory Dose

• Dose Limit #

Control Room Whole Body 0.1 0.4 5 Thyroid 16.2 29.2 30 Skin 4.8 12.5 30 EAB Whole Body 1.9 3.6 25 Thyroid 157.9 140.8 300 LPZ Whole Body 1.7 1.9 25 Thyroid 130.3 144.7 300  ;

l

' rounded to nearest tenth l

1. Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) dose limits are per 10CFR100.11. Control )

Room dose limits are per 10CFR50 Appendix A, General Design Criterion (GDC) 19 and NUREG 0800 Standard Review Plan (SRP) Section 6.4 As noted in the NEI Generic Framework Document (Reference 1), the acceptability of applications utilizing the revised accident source terms "may be judged by the same licensing acceptance limits (e.g., dose limits in 10CFR100) in use with the TID-14844 source term.

That is, the licensee would show that the revised design basis, with either selective or essentially complete application of NUREG-1465 together with the plant changes under evaluation, results in doses no greater than these licensing acceptance limits." The off-site

Attachment 1 PY-CEI/NRR-2076L Page 6 of 9 dose licensing acceptance limit for PNPP is 10CFR100.11 (see Question 3 of the Significant Hazards Consideration for details on the source of this PNPP licensing acceptance limit). As seen in the above Table, the newly calculated radiological doses are lower than the current analysis for six of the seven factors evaluated. For the one factor which was higher,i.e., at the EAB for thyroid dose (from 140.8 REM to 157.9 REM), the dose remained significantly below l the 10CFR100 limit of 300 REM to the thyroid. Consequently, the results of the LOCA reanalysis constitute a basis for demonstrating compliance with the requirements of 10CFR100 and with 10CFR50, Appendix A, GDC 19.

A significant amount ofinformation regarding the assumptions, conservatisms, and methodology of the dose calculations is provided in Attachment 6 (the non-proprietary dose calculations). The majority of that information is not repeated in this attachment. However, several points are emphasized in relation to the mitigation techniques employed in response to the postulated design basis Loss of Coolant Accident. As noted above, knowledge of the more physically correct source term timing and chemical form permits use of more appropriate mitigation techniques. Specifically, natural forces such as gravitational settling of aerosol (particulates) have been credited inside the drywell and in portions of the main steam lines, which significantly reduces the amount of radionuclides that could escape from the containment and into the emironment. Also, based on a high radiation signal in the Control Room, the Containment Spray system would be operated post-LOCA for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (previous analyses r.ssumed 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of spray operation), in order to scrub released radionuclides from the containment atmosphere and into the suppression pool, and thus reduce the post-LOCA off-site and Control Room dose. Once the containment sprays have been successful in sweeping the iodine to the suppression pool, the iodine must be retained in the water. To achieve this, the pH level of the suppression pool will now be raised to 7 or above following the accident, and then maintained at 7 or above. This prevents significant fractions of the dissolved iodine from being converted to elemental iodine and then re-evolving to the containment atmosphere. During the course of the accident the pH of the suppression pool can decrease due to radiolysis of reactor coolant and chloride-bearing electrical insulation, which would create acids. The method for pH control will use the existing Standby Liquid Control (SLC) system for raising (and maintaining )

long term post-accident suppression pool pH levels to 7 or above. Calculations have shown that the contents of one tank of the Standby Liquid Control solution will be effective in raising and maintaining pH levels for 30 days following the DBA. Controls over SLC system operability are already included in the Technical Specifications. In addition, a backup method for pH control will be developed for use if post-accident suppression pool sampling identifies that the primary pH control method is not being effective.

Necessary procedure changes and appropriate training will be made in conjunction with implementation of this change to reflect these revised mitigation techniques.

Post-accident operator actions are minimized. The operator action associated with initiating the Containment Spray system does not change. Containment Spray is initiated via a push button in the Control Room. The previously required manual initiation of the MSIV-LCS involved multiple operator actions to open and close numerous valves and start the blowers, which will no

Attachment 1 PY-CEI/NRR-2076L Page 7 of 9 longer be required. Replacing these actions, the new analysis assumes the operator closes the ,

Main Steam Stop Valves (which was previously one of the steps in manually initiating the MSIV-LCS ), and, based on post-accident pH samples of the suppression pool, initiates the ,

Standby Liquid Control system, which is accomplished via a push button in the Control Room.  !

These operator actions are less complex than those previously required, and minimize the j probability of an error.

The Control Room dose calculations were performed utilizing the PNPP-specific atmospheric dispersion factors (Chi /Q's) currently listed for the Control Room in the Updated Safety Analysis Report (USAR). These Chi /Q values were determined based on a special study i performed at PNPP (Reference 10), which is included for NRC staff review as Attachment 5 to this letter.

The new Main Steam Stop Valve specification 3.6.1.9 (see Attachment 3 for the marked-up pages) retains appropriate controls over the operability of these valves. Previously, operability of the Main Steam Stop Valves was an integral part of the MSIV-LCS system functional test (SR 3.6.1.9.3), since closure of the valves was necessary for the outboard MSIV-LCS to function. The new LCO requires the valves to be OPERABLE. The ACTIONS taken if a Main Steam Stop Valve is inoperable require the affected main steam line to be isolated in a fashion that would create a holdup volume for the leakage past the MSIVs. This must be performed within 30 days, or the plant is required to be shutdown. The 30 day Completion Time is consistent with the Completion Time that was provided for the MSIV-LCS. The new leakage control method (using the MSSVs) therefore uses that same duration. During this 30 day Completion Time, the remaining OPERABLE MSIVs in that main steam line are adequate to perform the required leakage holdup function. However, the overall reliability is reduced ,

because a single failure of an MSIV in that line could result in a loss of the MSIV leakage  !

holdup function. The purpose of closing two valves in a main steam line is based on the characteristics of the revised design basis accident source term (i.e., predominantly aerosol), and provides a holdup volume within the main steam line for deposition of the aerosol on the inner walls of the raain steam line. If an MSSV is " inoperable", but closed, credit can be taken for it ,

in meeting the ACTION. Leak tightness of the MSSVs is not necessary to ensure the assumptions of the dose calculation methodology are met for the main steam lines, since leakage flow characteristics used in the analyses are affected only by the turbulence caused by an open ended pipe (i.e., the Main Steam Stop Valves fail to close). The 30 day Completion Time is based on the redundant capability afforded by the remaining OPERABLE MSIVs and the low probability of a DBA LOCA occurring during this period. As long as two of the valves in that main steam line are closed, a holdup volume can be sustained and the plant can continue .

operation. The ACTIONS are modified by a Note allowing separate condition entry for each penetration flow path because an inoperable MSSV in a main steam line does not affect the ability to provide a holdup volume in the affected line (between the MSIVs) or in the other lines (between the MSIVs and/or the MSSVs). The Required Actions provide appropriate compensatory actions for each inoperable MSSV. Complying with the Required Actions may allow for continued operation, and subsequent inoperable MSSVs are governed by subsequent Condition entry and application of associated Required Actions.

Attachment 1 PY-CEI/NRR-2076L Page 8 of 9 REFERENCES

1. NEI document (prepared by the Electric Power Research Institute) entitled " Generic Framework for Application of Revised Accident Source Term to Operating Plants",

EPRI TR-105909, Interim Report, November 1995.

2. PSAT 04212H.03, " Ultimate Iodine Decontamination Factor for Perry DBA"

3. PSAT 04202H.04, " Aerosol Decay Rates (Lambda) in Drywell"

4. PSAT 04202H.05, " Aerosol Decay Rates (Lambdas) in Containment with Spray"

5. PSAT 04212H.06, " Mixing Between the Sprayed and Unsprayed Portions of the Perry Containment"

6. PSAT 04202H.08, "Steamline: Particulate Decontamination Calculation"

7. PSAT 04202H.09, " Steam Line: Elemental Iodine Decontamination Calculation"

8. PSAT 04202H.12, " Calculation of Fraction of Containment Aerosol Deposited in Water"

9. PSAT 04202H.13, "Off-site and Control Room Dose Calculation"

10. NUS-4792, "Results of the Atmospheric Tracer Study Within the Building Complex at the PNPP" COMMITMENTS WITHIN THIS LETTER Identified below are the actions committed to in this letter.

• Per agreements reached with the NRC through the BWROG program for increasing main steam line leakage rate limits, a commitment is made that if the leakage rate on a main steam line exceeds 100 scfh during Technical Specification required testing, the leakage rate for that line will be restored to s 25 scfh when tested at Pa (the current criterion for leakage),

prior to plant restart.

. Based on a high radiation signal in the Control Room, the Containment Spray system would be operated post-LOCA for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (previous analyses assumed 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of spray operation), in order to scrub released radionuclides from the containment atmosphere and into the suppression pool, and thus reduce the post-LOCA off-site and Control Room dose.

• The pH level of the suppression pool will be raised to 7 or above post-LOCA, and then maintained at 7 or above. The method for pH control will use the existing Standby Liquid Control (SLC) system.

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• A backup method for pH control will be developed for use if post-accident suppression pool sampling identifies that the primary pH control method is not being effective.

• Necessary procedure changes and appropriate training will be made in conjunction with l implementation of this change to reflect these revised mitigation techniques.

Attachment 1 PY-CEI/NRR-2076L Page 9 of 9 ENVIRONMENTAL CONSIDERATION The proposed Technical Specification change request was evaluated against the criteria of 10CFR51.22 for environmental considerations. The proposed change does not significantly increase individual or cumulative occupational radiation exposures, does not significantly change the types or significantly increase the amounts of effluents that may be released off-site and, as discussed in Attachment 2, does not involve a significant hazards consideration. Based on the foregoing, it has been concluded that the proposed Technical Specification change meets the criteria given in 10CFR51.22 (c)(9) for categorical exclusion from the requirement for an EnvironmentalImpact Statement.

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Attachment 2 SIGNIFICANT HAZARDS CONSIDERATION PY-CEI/NRR-2076L Page 1 of 6 The standards used to anive at a determination that a request for amendment involves no significant hazards considerations are included in the Commission's Regulations,10CFR50.92, which state that the operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any previously evaluated; or (3) involve a significant reduction in a margin of safety.

The proposed amendment has been reviewed with respect to these three factors and it has been determined that the proposed change does not involve a significant hazard because:

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change removes the Teciaical Specification requirements for the Main Steam Isolation Valve Leakage Control t ystem (MSIV-LCS), and increases the allowable leak rate specified for the main steam line.s. Although the requirements for the MSIV-LCS are being removed (since credit is no longer taken for the system as part of the design basis accident analysis), OPERABILITY requirements on the Main Steam Stop Valves are being retained since the valves meet Criterion 3 of10CFR50.36(c)(2)(ii). Removing the Technical Specification requirements of the MSIV-LCS and increasing main steam line allowable leakage rates has been addressed in the Loss of Coolant Accident (LOCA) reanalysis and does not adversely affect operation of other equipment or systems important to safety. These changes do not affect the precursors for accidents or transients analyzed in Chapter 15 of the Perry Nuclear Power Plant (PNPP) Updated Safety Analysis Report (USAR). Therefore, there is no increase in the probability of accidents previously evaluated.

The spectrum of LOCAs was considered to determine which would be most limiting with respect to radiological consequences. The worst case LOCA (i.e., main steam line break upstream of the inboard MSIV) off-site and Control Room doses have been reanalyzed using the revised design basis accident (DBA) source term (from NUREG-1465 and the Nuclear Energy Institute (NEI) document " Generic Framework for Application of Revised Accident Source Term to Operating Plants") in order to assess the radiological consequences of the increased main steam line leak rates, and not taking credit for the MSIV-LCS. The radiological analysis used conservative assumptions and analytical l techniques. These conservatisms in the LOCA reanalysis have been determined to be comparable to the conses catisms utilized in the original analyses.

Attachment 2 SIGNIFICANT HAZARDS CONSIDERATION PY-CEI/NRR-2076L '

Page 2 of 6 The results of the off-site and Control Room dose reanalysis are provided below.

DOSE RESULTS (REM)

Proposed Existing USAR USAR Regulatory Dose

• Dose Limit #

l Control Room Whole Body 0.1 0.4 5 I Thyroid 16.2 29.2 30 Skin 4.8 12.5 30 l

EAB Whole Body 1.9 3.6 25 Thyroid 157.9 140.8 300 f LPZ Whole Body 1.7 1.9 25 Thyroid 130.3 144.7 300 4

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• rounded to nearest tenth i # Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) dose limits are per 10CFR100.11. Control Room dose limits are per 10CFR50 App. A, General Design Criterion (GDC) 19 and NUREG 0800 Standard Review Plan (SRP) Section 6.4

As noted in the NEl Generic Framework Document (" Generic Framework for Application

of Revised Accident Source Term to Operating Plants", EPRI TR-105909, Interim Report, November 1995), the acceptability of applications utilizing the revised accident source terms "may be judged by the same licensing acceptance limits (e.g., dose limits in 10CFR100) in use with the TID-14844 source term. That is, the licensee would show that i the revised design basis, with either selective or essentially complete application of NUREG-1465 together with the plant changes under evaluation, results in doses no greater than these licensing acceptance limits." The off-site dose licensing acceptance limit for PNPP is 10CFR100.11 (see Question 3 for details on the source of this PNPP licensing

! acceptance limit). The newly calculated radiological doses were lower for six of the seven factors evaluated. For the one factor which was higher, i.e., at the EAB for thyroid dose

! (from 140.8 REM to 157.9 REM), the dose remained significantly below the 10CFR100

) limit of 300 REM to the thyroid. This analysis demonstrated that the resulting off-site and Control Room doses were well below the regulatory limits contained in 10CFR100, Reactor Site Criteria, and 10CFR50, Appendix A, General Design Criterion 19, Control Room. Therefore, the proposed changes do not involve a significant increase in the consequences of previously evaluated accidents.

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Attachment 2 SIGNIFICANT HAZARDS CONSIDERATION PY-CEI/NRR-2076L Page 3 of 6 )

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2. The proposed change would not create the possibility of a new or different kind of accident from any previously evaluated.

The proposed change removes the Technical Specification requirements for the MSIV-LCS, retains the Technical Specification requirements for the Main Steam Stop Valves, and increases the allowable leak rate specified for the main steam lines.

Removing the Technical Specification requirements for the MSIV-LCS is based on reanalysis of off-site and Control Room doses, where the MSIV-LCS is not credited in the calculation. As noted above, the reanalysis utilizes the revised design basis accident (DBA)

source terms. The limiting reanalysis case assumes that main steam line leakage is attenuated in the main steam line from the reactor vessel out to the outboard MSIV. This is the limiting scenario since the worst case single failure, and hence the most limiting analysis case, involves a failure to close the valve downstream of the outboard MSIV in each main steam line, i.e., the Main Steam Stop Valves (INIIF0020A,B,C and D). Although this most limiting analysis case assumes a failure to close the Main Steam Stop Valves, retention of OPERABILITY requirements on these valves is appropriate to ensure the single failure analysis ren,ains valid.

Not crediting the MSIV-LCS in the design basis accident analysis is consistent with the approach taken by several BWR licensees, which have applied for NRC approval of this change using an approach developed by the Boiling Water Reactor Owners Group (BWROG). The BWROG methodology involves seismically qualifying the main steam lines out to and including the non-safety related, non-seismic drain line and main condenser, and then using that volume to attenuate leakage past the MSIVs. At PNPP, the existence of safety related, seismically qualified piping leading to the safety related, Class lE powered Main Steam Stop Valves (downstream of the outboard MSIV), together with the characteristics of the revised accident source term (i.e., predominantly aerosol which is largely retained in the drywell, containment and main steam lines) provides the option of taking credit only for the volume within the main steun lines for leakage attenuation.

Knowledge of the more physically correct source term timing and chemical form permits use of more appropriate mitigation techniques. Specifically, natural forces such as gravitational settling of aerosol (particulates) has been credited inside the drywell and in portions of the main steam lines, which significantly reduces the amount of radionuclides that could escape from the containment and into the environment. Also, based on a high radiation signal in the Control Room, the Containment Spray system would be operated post-LOCA for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (previous analyses assumed 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of spray operation), in order to scrub released l

radionuclides from the containment atmosphere and into the suppression pool, and thus reduce the post-LOCA off-site and Control Room dose. Once the containment sprays have been successful in sweeping the iodine to the suppression pool, the iodine must be retained

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Attachment 2 SIGNIFICANT HAZARDS CONSIDERATION PY-CEI/NRR-2076L Page 4 of 6 in the water. To achieve this, the pH level of the suppression pool will now be raised to 7 or above following the accident, and then maintained at 7 or above. This prevents significant fractions of the dissolved iodine from being converted to elemental iodine and then re-j evolving to the containment atmosphere. During the course of the accident the pH of the j suppression pool can decrease due to radiolysis of reactor coolant and chloride-bearing l electrical insulation, which would create acids. The method for pH control will use the j existing Standby Liquid Control (SLC) system for raising (and maintaining ) long term post-accident pH levels to 7 or above. Calculations have shown that the contents ofone tank of the Standby Liquid Control solution will be effective in raising and maintaining pH levels for 30 days following the DBA.

Post-accident operator actions are minimized. The operator action associated with initiating the Containment Spray system does not change. Containment Spray is initiated sia a push button in the Control Room. The previously required manual initiation of the MSIV-LCS involved multiple operator actions to open and close numerous valves and start the blowers, which will no longer be required. Replacing these actions, the new analysis simply assumes the operator closes the Main Steam Stop Valves (which was previously one of the steps in manually initiating the MSIV-LCS system), and, based on post-accident pH samples of the suppression pool, initiates the Standby Liquid Control sytem, which is accomplished via a push button in the Control Room. These operator actions are less complex than those previously required, and minimize the probability of an error.

Other accidents, as described in USAR section 15, were reviewed. The original methodology, input parameters and overall conclusions contained within these accident evaluations were found to be unaffected by the changes proposed by this activity.  ;

Removing the Technical Specification requirements of the MSIV-LCS and increasing MSIV allowable leakage rates has been addressed in the LOCA reanalysis and does not adversely affect operation of other equipment or systems important to safety. This activity does not alter or impact plant systems, structures or components which were not appropriately addressed in the LOCA reanalysis. No new accident initiator or failure mode is introduced. The physical isolation of the MSIV-LCS from the Main Steam system will eliminate leakage pathways. This modification will be performed as part of the PNPP design change process.

With respect to the change in main steam line leakage limits, the BWROG has concluded, based on an in-depth evaluation of MSIV leakage (as discussed in NEDC-31858P "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems", Revision 2, and summarized in NUREG-1169 " Technical Findings

! Related to Generic Issue C-8; Boiling Water Reactor Main Steam Isolation Valve and l Leakage Treatment Methods"), that leakage rates of up to 500 scfh are not indicative of 1

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i Attachment 2 SIGNIFICANT HAZARDS CONSIDERATION PY-CEl/NRR-2076L Page 5 of 6 1

I substantial mechanical defects in the valves which would challenge the capability of the valves to fulfill their safety function ofisolating the steam lines. Therefore, as demonstrated l in the design basis LOCA radiological reanalysis, the proposed increased allowable MSIV leakage rate (i.e., each line less than or equal to 100 scfh and total leakage less than or equal i to 250 scfh when tested at Pa) will not affect each MSIV's isolation function capability.

l Additionally, no new operator actions or errors are introduced as a result of the increased main steam line leakage limits, other than those addressed above.

Based on the above discussions, the proposed change would not create the possibility of a new or different kind of accident from any previously evaluated.

3. The proposed change will not involve a significant reduction in the margin of safety.

The worst case LOCA (i.e., a main steam line break upstream of the inboard MSIV) has been reanalyzed using the revised DBA source term (NUREG-1465 and the NEI generic j framework document) in order to assess the radiological consequences of the increased j MSIV leak rates, and not taking credit for MSIV-LCS. The radiological analyses used l l

conservative assumptions and analytical techniques. The results of the revised DBA source term dose calculations should be determined acceptable using the current licensing basis acceptance limits (those that were used for initial plant licensing).

As noted in the NEI Generic Framework Document (" Generic Framework for Application of Revised Accident Source Term to Operating Plants", EPRI TR-105909, Interim Report, November 1995.), "to demonstrate that an adequate margin of safety is maintained, the licensee may show that the doses associated with the revised design basis (resulting from the revised source term together with the plant change under evaluation) are less than the licensing acceptance limits for the plant."

The licensing acceptance limits for off-site dose are discussed in Supplement 8 to the NRC l Safety Evaluation Report (SER) for PNPP, Section 15.3 " Radiological Consequences of j Design Basis Accidents". The licensing acceptance limits are the guideline values of i 10CFR100.11," Reactor Site Criteria". The SER states "The doses computed for this accident are less than the guideline values of 10CFR100.11 and the staff concludes that the Perry plant is adequately designed to mitigate the off-site consequences arising from a LOCA." For Control Room doses, the licensing acceptance limit is discussed in J Supplement 10 to the NRC SER, Section 6.4 " Control Room Habitability". The licensing acceptance limits are as stated therein, i.e., "The staffs LOCA analysis indicates that the Control Room doses are within the guidelines of General Design Criterion (GDC) 19 of Appendix A to 10CFR50 and of Section 6.4 of the Standard Review Plan (SRP, NUREG-0800)."

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I Attachment 2 SIGNIFICANT HAZARDS CONSIDERATION PY-CEI/NRR-2076L -

Page 6 of 6 t i

l The revised PNPP design basis calculations (i.e., the revised DBA source term coupled ,

with the plant changes under evaluation) demonstrated that the resulting efr-site and i l

Control Room doses were below the licensing acceptance limits contained in 10CFR100, 10CFR50, Appendix A, General Design Criterion 19, and SRP Seciian 6.4. An acceptable ,

margin of safety is inherent in these licensing acceptance limits. The improvement in the L technical knowledge base and in the analytical techniques that are part of the revised accident source term, and the modeling of the increased MSIV leakages without taking credit for MSIV-LCS, do not alter the acceptability of the margin. Therefore, the resulting i calculated Control Room and off-site doses, w%h are well within regulatory limits, ensure that the proposed change does not invoht. a significant reduction in the margin of safety.

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