View source for NG-94-2629, Application for Amend to License DPR-49,proposing Changes to Ts,By Increasing Allowable Main Steam Isolation Valve Leakage & Deleting TS Requirements Applicable to MSIV LCS

{{Adams
| number = ML20072F161
| issue date = 08/15/1994
| title = Application for Amend to License DPR-49,proposing Changes to Ts,By Increasing Allowable Main Steam Isolation Valve Leakage & Deleting TS Requirements Applicable to MSIV LCS
| author name = Franz J
| author affiliation = IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT
| addressee name = Russell W
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| docket = 05000331
| license number = 
| contact person = 
| document report number = NG-94-2629, NUDOCS 9408230232
| package number = ML20072F163
| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS &amp; PERMITS
| page count = 15
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:Ilifs UTILITIES INC.
John F. Franz, Jr.
Vice President, Nuclear August 15, 1994 NG-94-2629 Mr. William T. Russell, Director Office of Nuclear Reactcr Regulation U. S. Nuclear Regulatory Commission Attn:   Document Control Desk Mail Station Pl-137 Washington, DC 20555

==Subject:==
Duane Arnold Energy Center Docket No: 50-331 Op. License No: DPR-49 Request for Technical Specification Change (RTS-232):              Increase in Allowable MSIV Leakage Rate and Deletion of the MSIV Leakage Control System

==Reference:==
NEDC-31858P, Rev. 2:                 "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems" 4                               File:        A-117, N-11

==Dear Mr. Russell:==

In accordance with the Code of Federal Regulations, Title 10, Nu      _ Sections 50.90 and 50.59, IES Utilities Inc. hereby requests
   ,.    : revision of the Duane Arnold Energy Center (DAEC) Technical Specifications (TS), Appendix A to Operating License DPR-49.
The proposed changes increase the allowable main' steam isolation valve-(MSIV) leakage and delete the TS requirements applicable to the MSIV~ leakage control system (LCS).                         IES Utilities Inc.
proposes to eliminate the LCS and instead direct MSIV leakage to the main steam drain lines and the main condenser.                                It has been demonstrated that this is a more reliable method than the LCS for treating MSIV leakage;and it presents no undue risk to public health and safety. While certain main steam piping and components which are used in this method are not classified as Seismic Category I, a detailed evaluation indicates that the affected portions of the main steam system piping and equipment are seismically rugged and satisfy the intent of Appendix A of 10 CFR 100 for seismic adequacy.
.        These; proposed TS changes are supported by work performed by the Boiling ~ Water Reactor Owners' Group (BWROG) with IES Utilities Inc. participation. This work is documented in General Electric Company. Report, NEDC-31858P, Rev. 2, entitled "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage
        ~enntrol Systems."         Our submittal is consistent with that of the 9408230232 940015 PDR  ADOCK 05000331 P                  PDR 0p{
General Office
* P.O. Box 351
* Cedar Rapids, Iowa 52406
* 319/398-4411         1 An IES INDUSTRIES Company j
                                                                                                        }

y -                                      _                                                ,
Mr. William T. Russell NG-94-2629 August 15, 1994 Page 2 Edwin I. Hatch Nuclear Plant, Unit 2 (Docket No. 50-366), for which the NRC issued a Safety Evaluation and TS Amendment dated March 17, 1994.
Attachment 1 provides a summary of the methodology used to develop this submittal. Attachment 2 discusses the basis for the change. Attachment 3 provides marked-up copies of the current TS pages. Attachment 4 details the basis for IES Utilities Inc.'s determination that the proposed changes do not involve a significant hazards consideration. Attachments 5 and 6 provide the safety ansessment and discussion of environmental consideration (respectively). Attachment 7 contains the seismic evaluation for the DAEC.
We plan to implement these changes during the next refueling outage, currently scheduled to begin February 24, 1995.
Consequently, IES Utilities Inc. requests approval of the proposed amendment by February 1, 1995.
This application has been reviewed by the DAEC Operations Committee and the DAEC Safety Committee.            Pursuant to the requirements of 10 CFR 50.91, a copy of this submittal, including the no significant hazards consideration analysis, is being forwarded to our appointed state official.
Should you have any questions regarding this matter, please contact this office.
This letter is true and accurate to the best of mf knowledge and belief.
IES Utilities Inc.
By                  /   . _ 1/M John F.(Frank Vi/e President, Nuclear
    -State of Iowa (County) of Linn Signed and sworn to before me on this          l 'f    day of       /9(N-US I      ,
1994, by     JcH d f' .1'AA m..              .
                                                                )
( b1Tb (r        /     ff!nfhrQse
                                                                           ^
Notary / Public in and          or the(ptate of Iowa In N T /t YI    19 / (-
Commission Expirds JFF/CJR/pjv

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Mr. William T. Russell NG-94-2629 August 15, 1994                                                     !
Page 3                                                              l l
Attachments:     1)  Summary of Methodology and Evaluations        1
: 2)  Basis for Change Request
: 3)  Proposed Change (RTS-232) to the Duane Arnold Energy Center Technical Specifications
: 4)  Evaluation of Change with Respect to 10 CFR 50.92
: 5)  Safety Assessment
: 6)  Environmental Consideration
: 7)  Seismic Evaluation cc:  C. Rushworth L. Liu L. Root R. Pulsifer (NRC-NRR)
J. Martin (Region III)
S. Brown (State of Iowa)                                      i NRC Resident Office DCRC l
1
                                                                      )
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RTS-232                                                                                                                       Attachment 1 to NG-94-2629 Page 1 of 3 Summary of Methodology and Evaluations In 1986, the Boiling Water Reactor Owners' Group (BWROG) formed the Main Steam Isolation Valve (MSIV) Leakage Closure Committee to resolve the issue of MSIV leakage. The BWROG Committee studied the issues of MSIV leakage rates and associated excessive maintenance required for the MSIVs and the leakage control                                                                                  i systems (LCS). As a resolution of those issues, the BWROG                                                                                  i proposed to use the main steam piping, drain lines, and isolated condenser as an alternate method for MSIV leakage treatment.                                                                                 ,
This alternate MSIV leakage treatment method has been shown to                                                                               I I
provide effective and reliable fission product attenuation for reducing the radiological consequences of MSIV leakage. This leakage treatment method takes advantage of the large volume in                                                                            j the main condenser to provide hold-up and plate-out of fission                                                                              i products that may leak from closed MSIVs.
A plant-specific radiological dose calculation has been performed which showed that for MSIV leakage at the DAEC of 100 scfh per MSIV with a maximum total leakage of 200 scfh, the loss of coolant accident (LOCA) doses would remain within the regulatory guidelines. The radiological dose calculations used the methodology developed by General Electric for the BWROG. This methodology is documented in Appendix C of NEDC-31858P, Revision 2, "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems."
l                                 In order to justify the capability of the mal                                                                           steam piping and I
condenser as the alternate leakage treatment system, IES Utilities Inc. has verified that the main steam lines, the main steam drain line, the condenser, and interconnecting piping and equipment are seismically adequate to withstand a safe shutdown earthquake and maintain their integrity. The seismic adequacy of these piping and equipment systems at the DAEC was confirmed by comparing them to a detailed earthquake experience database as discussed in Section 6.7 of NEDC-31858P, Revision 2, and performing engineering walkdowns and evaluations using seismic capability engineers.
The earthquake experience database, which consists of the documentation of the performance of piping and equipment in power and industrial facilities during past earthquakes, is founded on extensive studies of over 100 industrial facilities and surveys of several hundred other facilities located in the vicinity of numerous strong motion earthquakes that have occurred in California, Alaska, New Zealand, and Latin American countries since 1971.
l _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _                                     -

RTS-232                                              Attachment 1 to NG-94-2629 Page 2 of 3 The current standard practice for the seismic dasign of piping and equipment systems has not considered the real performance of such systems in strong motion earthquakes. This has resulted in extreme conservatism in the treatment of primary stresses when uncorrected linear elastic analyses are performed and the results are compared to stress limits based on static tests. The earthquake experience data provides the only available full-scale tests of designs and installations. The data, therefore, provides realistic and practical criteria for determining the seismic adequacy of piping and equipment.
Equipment and above ground piping at database facilities have exhibited excellent resistance to damage during and after earthquakes without the specific application of seismic design considerations and provisions. A large number of classes of equipment (pumps, valves, tanks, instrument cabinets, etc.) have proven seismically rugged when properly anchored. For welded steel piping designed and constructed to normal industrial practice (e.g., ANSI B31.1), past seismic experience has never shown a primary collapse mode of failure. A relatively small number of seismically induced piping failures have occurred due to excessive relative support movements or seismic interactions.
Consistent with the verification methodology, a plant specific seismic verification walkdown of all systems and components associated with the alternate MSIV leakage treatment was performed by seismic capability engineers. The purpose of the walkdown was to physically verify that the components in the alternate leakage treatment system have attributes similar to those in the database that have demonstrated good seismic performance in past earthquakes and to identify potential seismic vulnerabilities. As a result of the walkdown and subsequent evaluations, IES Utilities Inc. has determined that the plant features compare well with the database.
The walkdown also included an inspection for those structural details and causal factors that resulted in component damage at industrial sites contained in the database to ensure that such conditions are evaluated to satisfaction or that plant modifications are implemented to resolve the concern. An engineering analysis of selected critical supports was performed which showed that the supports exhibited substantial margin. As a result of the walkdown, IES Utilities Inc. Identified the need to implement minor modifications or repairs to several components. These components and the required modifications or repairs are discussed in Attachment 7. The appropriate modifications or repairs will be completed prior to implementing the proposed TS change. Taking the measures outlined in Attachment 7 ensures resolution of all of the identified outliers and ensures that the damage reported for the database components

RTS-232                                             Attachment 1 to NG-94-2629 Page 3 of 3 should not occur to the DAEC main steam piping and condenser or to the associated support systems.
In summary, IES Utilities Inc. has concluded that the DAEC main steam lines, main steam drain lines, condenser, and applicable interconnecting piping and equipment, are well represented by the earthquake experience data demonstrating good seismic performance, are confirmed to exhibit excellent resistance to damage from a design basis earthquake, have been shown to have substantial margin for seismic capability, and are, therefore, seismically adequate to withstand the DAEC design basis earthquake and maintain pressure retaining integrity. This capability of the alternate MSIV leakage treatment system to withstand the effects of the safe shutdown earthquake and continue to perform its intended function (treatment of MSIV leakage) satisfies the intent of the seismic requirement of Appendix A to 10 CFR 100.

RTS-232                                               Attachment 2 to NG-94-2629 Page 1 of 9 l
l Basis for Change Request                      I Proposed Change 1 - MSIV Leakage Limit Increase l
This proposed change increases the allowable leak rate specified in Technical Specifications (TS) 4.7.A.2.c.3 from 11.5 standard cubic feet per hour (scfh) for any one main steam isolation valve (MSIV) when tested at 24 psig to 100 scfh for any one MSIV with a total maximum pathway leakage rate of 200 scfh through all four main steam lines when tested at 24 psig. If an MSIV exceeds 100 scfh, it will be restored to less than or equal to 11.5 scfh.
Basis for Proposed Change 1 The current TS allowable MSIV leakage rate is extremely limiting and routinely requires repair and retest of the MSIVs. This can significantly impact the maintenance work load during plant outages and contribute to outage extensions. The outage planning group at the DAEC typically schedules several days of contingency time for repair and retest of the MSIVs. The proposed increase in the allowable MSIV leakage rate would reduce the need for repair and, thereby, reduce dose to maintenance personnel consistent with As Low As Reasonably Achievable (ALARA) principles.
Failures of MSIVs to meet the current typical TS leakage limit have been documented in response to surveys conducted by the NRC during the early 1980s and by the Boiling Water Reactor Owners' Group (BWROG) during the middle and late 1980s. As many as 50 percent of the total "as found" MSIV local leak rate test (LLRT) results were reported in the early NRC survey to exceed the leakage rate limit.
The BWROG, with IES Utilities Inc. participation, studied the issues regarding MSIV leakage rates, their causes, and available alternatives. The results of the BWROG study are provided in NEDC-31858P, "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems," Revision 2 and are also summarized in NUREG-1169. In response to Generic Issue C-8, "MSIV Leakage and LCS Failure," the BWROG has recommended corrective actions and maintenance practices to reduce the MSIV leakage rates.
A survey conducted by the BWROG of MSIV LLRT results between 1984 and 1988 indicated that the implementation of industry and BWROG actions has been effective in reducing the MSIV leakage rates.
Of key importance was the reduction in the number of valves which experienced substantially high leakage rates. However, the survey also concluded that about 23 percent of the total "as

I l
RTS-232                                                 Attachment 2  l to NG-94-2629 Page 2 of 9 1
l 1
found" MSIV leakage rates still exceeded the limit of 11.5 scfh and about 10 percent exceeded 100 scfh.
Despite the improvement in leakage performance, MSIV leakage rates still frequently exceed the current typical TS limit and the resultant maintenance problems, although less severe, remain as a significant issue. Furthermore, based on extensive evaluation of valve leakage data, the BWROG has found disassembling and refurbishing the MSIVs to meet very low leakage limits frequently contributes to repeating failures. In most cases, machining of the valve seat is required to meet the current TS limits. Each time the seat is machined, the thickness is reduced, leading to earlier than necessary seat replacement.
Disassembly and assembly also cause wear on the various components removed and replaced. By not requiring disassembly and refurbishment of the valves for minor leakage, the utility reduces its susceptibility to one of the contributors to recurring valve leakage problems which lead to later LLRT failures and the possibility of comprr?ising plant safety.
The current TS allowable MSIV leakage rate (11.5 scfh) is extremely conservative considering the valve's physical size and operating characteristics (large size and fast-acting).
Additionally, the ability of the turbine building equipment to contain the radioactive material was not considered at the time the leakage limit was established. Based on the in-depth evaluation of MSIV leakages, 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 capability of the valves to fulfill their safety function of isolating the steam lines. Furthermore, valve manufacturers have stated that leakage rates of up to 200 scfh can occur without having a major valve defect. Therefore, the proposed increase from 11.5 scfh per MSIV to 100 scfh per MSIV will not affect the MSIV's isolation function performance. Additionally, processing the post LOCA releases through the steam line drains and the condenser is highly effective, resulting in no significant impact on the health and safety of the public.
This proposed increase in the allowable MSIV leakage rate             ;
provides a more realistic, but still conservative, limit for the MSIVs. This increase in allowable leakage will result in more successful local leak rate tests, which in turn will result in significantly reduced MSIV maintenance costs, reduced dose exposure to maintenance personnel, reduced outage durations and       j extend the effective service life of the MSIVs.                       1 1
The radiological dose methodology developed by General Electric       1 (GE) for the BWROG was used to calculate the effects of the           )

RTS-232                                              Attachment 2 to NG-94-2629 Page 3 of 9 increased MSIV leakage rate. New control room, technical support center (TSC) and offsite doses have been calculated for a postulated loss of coolant accident (LOCA). These new doses are prestoted in Table 1. The analyses demonstrate that an MSIV leakag- rate of 100 scfh per MSIV (200 scfh total for 4 main steam lines (MSLs)) results in acceptable dose exposures for the control room, TSC, and offsite boundaries.
Proposed Change 2 - MSIV LCS Elimination This proposed change deletes TS 3.7.E and 4.7.E and Bases section 3.7.E and 4.7.E, thus eliminating the MSIV leakage control system (LCS) requirements from the TS.
Basis for Proposed Change 2 In 1983, Generic Issue C-8 was established to track the resolution of an NRC concern that MSIV leakage rates, as determined by conservative local leak rate tests, were too high and the LCS would not function at high MSIV leakage rates. A 1981 NRC survey of the industry indicated that 33 percent of the total "as found" LLRT conditions for MSIVs exceeded leakage rates of 100 scfh. Since the process capability of the LCS at the DAEC is designed for MSIV leakage rates of no more than 100 scfh, the potential exists for the LCS not to function as analyzed for a design basis LOCA.
IES Utilities Inc. proposes to delete the LCS requirements from the TS and to use the main steam drain lines and the isolated main condenser as an alternate method for MSIV leakage treatment.
In addition to resolving the concern identified in Generic Issue C-8, the proposed deletion of the LCS requirements from the TS will result in significant operational and maintenance benefits.
LCS equipment is located in a high temperature, high radiation area and is required to be environmentally qualified, necessitating extensive preventive maintenance. The system has extensive logic and instrumentation which requires frequent calibration to meet the TS requirements. The BWROG has evaluated recent LCS performance data and the results are shown in NEDC-31858P, Rev. 2    The evaluation indicates that the LCS is extremely difficult to maintain, and as a result of maintenance requirements, plant shutdowns and startup delays have occurred within the industry.
The BWROG, with IES Utilities Inc. participation, has evaluated several alternate MSIV leakage treatment methods and has recommended the isolated condenser for MSIV leakage treatment.
This leakage treatment method takes advantage of the large volume in the isolated main condenser to hold up the release of any

RTS-232                                               Attachment 2 to NG-94-2629 Page 4 of 9 fission products potentially leaking from the closed MSIVs. The main steam drain lines are employed to convey leakage to the condenser. Since simpler and less equipment is employed, the alternate method is more reliable than the LCS. As supported by the BWROG, this proposed change will resolve the concern associated with LCS performance capability at high MSIV leakage rates, and will assure that a reliable and effective method is available for treating any potential MSIV leakage during a postulated LOCA. IES Utilities Inc. will incorporate the applicable alternate leakage treatment methods into DAEC procedures as appropriate.
Figure 1.1 of this Attachment shows the proposed primary and alternate drain paths. The proposed primary drain path at DAEC employs an MSL drain downstream of the MSIVs. There are two motor-operated valves (MOVs) in series in this line between the MSLs and the main condenser. Both valves must be open to establish the required drain path. Both MOVs will be provided with essential power to assure that they can be opened following the DBA LOCA to establish a large enough drain path to support the radiological analysis.
An alternate drain path will be available to convey MSIV leakage to the isolated condenser if either MOV fails to open. The alternate drain path consists of the bypass lines around the MOVs in the primary drain path. This alternate path contains a " fail open" valve and a restricting orifice. Consequently, if either primary MOV failed to open as required, the second drain path would be available to convey MSIV leakage to the main condenser.
Radiological dose calculations have been performed for this alternate path as well as for the primary path; dose exposures remain within the requirements of 10 CFR Part 100 for offsite doses and 10 CFR Part 50, Appendix A for the control room and       ;
TSC.                                                                J l
The BWROG has evaluated the availability of the main steam system   ;
piping and condenser as an alternate treatment pathway for          !
processing MSIV leakage. It was determined that the probability   j of a near coincident LOCA and seismic event is much smaller than other plant safety risks. The BWROG also determined that main steam piping and condenser designs exhibit substantial ruggedness, and that the ANSI B31.1 piping design requirements      i typically used for nuclear plant system designs contain             I significant design margin.
The BWROG contracted with EQE Engineering Consultants (EQE) to t perform a review of the performance of main steam system piping l and condensers in conventional power plants during past strong l
motion earthquakes. The review was based on the extensive

l RTS-232                                              Attachment 2 to NG-94-2629 i Page 5 of 9   ,
I earthquake experience database assembled by EQE for resolution of NRC Unresolved Safety Issue A-46. The EQE database includes hundreds of structures with a wide diversity of structural types and design criteria that house thousands of pipe runs, cable       1 trays, conduits, and related components. Thousands of equipment installations, from 1930s vintages to new items, are in the database. The 18 strong-motion earthquakes range in Richter magnitude from 5.4 to 8.1. EQE also obtained additional data for a wide range of both large- and small-bore piping which demonstrated good seismic performance during other strong motion earthquakes not covered in BWROG report NEDC-31858P, Rev. 2.
EQE also compared these piping systems and condensers to the piping systems and condensers typically used in domestic BWR plants. The result of the comparison supports the BWROG view that the main steam piping and condensers employed in GE BWRs      i would maintain their pressure-retention capability during a        ;
design-basis earthquake.                                           !
IES Utilities Inc. contracted with EQE to confirm the seismic capability of the DAEC's main steam piping and condenser to serve  l as an alternate leakage treatment system. Seismic verification     )
walkdowns were performed to assure that the MSLs, the main steam   ;
drain l'nes, the condenser, and interconnecting piping and equipment that were not seismically analyzed fall within the       ,
bounds of the design characteristics of the seismic experience     j database as discussed in Section 6.7 of the BWROG report. The    l walkdowns were conducted by seismic review teams consisting of at  i least two seismic capability engineers, one of whom was a          I licensed professional engineer. As a group, eacn team's members possessed a knowledge of the failure modes and performance of structures and piping in fossil fuel power plants during strong earthquakes, knowledge of nuclear design standards and seismic design practices for nuclear power plants and general knowledge    ,
of the plant systems functions.                                    i I
Those portions of the DAEC main steam and drain line/ bypass        l piping that were not seismically analyzed as part of the original   l plant design were evaluated to demonstrate that the piping and supports fall within the bounds of design characteristics found in conventional power plant steam piping which demonstrated good seismic performance. These conventional designs were shown to be comparable to the steam piping design for DAEC.
Portions of the main steam piping that were seismically analyzed as part of the original plant design included the main steam lines (from the outboard MSIV to the turbine stop valve), the main steam bypass line (to the bypass valves), and portions of various main steam branch connections to the seismic anchor

RTS-232                                                                            Attachment 2 to NG-94-2629 Page 6 of 9 downstream of the isolation valves for the branch. Design methods for these analyzed lines were consistent with seismic Category I qualification methods for DAEC.
The DAEC main condenser design is typical of those at facilities in the earthquake experience database that have experienced earthquakes in excess of the DAEC design basis earthquake. The DAEC condenser anchorage is comparable to the anchorage of earthquake experience database condensers.
The turbine building is classified as nonseismic, however the criteria for seismic Category I structures were used for the structural design of the entire building. A complete dynamic analysis was conducted for the turbine building. The same design procedures used for the reactor building were also used for the turbine building.                   Thus, the turbine building was specifically designed for seismic loading.
The DAEC main steam lines, main steam drain lines, condenser, and applicable interconnecting piping and equipment, are well represented by the earthquake experience data demonstrating good seismic performance, are confirmed to exhibit excellent resistance to damage from a design basis earthquake and have been shown to have substantial margin for seismic capability.
Therefore they are seismically adequate to withstand the DAEC design basis earthquake and maintain pressure retaining integrity. This capability of the alternate MSIV leakage treatment system to withstand the effects of the safe shutdown earthquake and continue to perform its intended function (treatment of MSIV leakage) satisfies the intent of the seismic requirement of Appendix A to 10 CFR 100.                                                            ,  contains the seismic evaluation which includes a complete list of the outliers identified during the walkdowns.
These outliers have been either evaluated or analyzed to demonstrate their acceptability as they currently exist, or plant modifications will be implemented to resolve the concerns. These components and the required modifications or repairs are discussed in Attachment 7. The appropriate modifications or repairs will be completed prior to implementing the proposed TS change. By taking the proposed measures to ensure resolution for all of the identified outliers, IES Utilities Inc. is assured that the damage reported for the database components should not occur to the DAEC main steam piping and condenser or to the associated support systems. Therefore, the proposed method for MSIV leakage treatment is seismically adequate and will serve as an acceptable alternative to the currently installed LCS.
The radiological dose methodology developed by GE for the BWROG

RTS-232                                                   Attachment 2 to NG-94-2629 Page 7 of 9 is documented in Appendix C of NEDC-31858P, Revision 2. This radiological analysis was used to calculate the effects of the proposed allowable MSIV leakage rate in terms of control room and       j offsite doses. While noc previously considered a requirement for the design of the LCS, dose calculations were also performed for the Technical Support Center (TSC). Table 1 shows the calculated dose exposures from the BWROG radiological analysis for the DAEC.
Regulatory limits and calculated doses from LOCA radiological analysis are also included in Table 1 for comparison purposes.
This analysis demonstrates that a leakage rate of 100 scfh per MSIV, with a maximum leakage of 200 scfh for all four MSLs (with the elimination of the LCS) results in an acceptable increase in the dose exposures previously calculated for the control room, TSC, exclusion area boundary (EAB), and low population zone (LPZ). The revised LOCA doses remain within the guidelines of 10 CFR 100 for offsite doses and 10 CFR 50, Appendix A, (General Design Criterion 19) for the control room and TSC doses.
Removal of the LCS will eliminate the system's impact on refueling and maintenance outage activities at the DAEC. The proposed alternate method (main steam lines and condenser) for MSIV leakage treatment will also eliminate the concern regarding LCS effectiveness at higher MSIV leakage rates.
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RTS-232                                                       Attachment 2 to NG-94-2629 Page 8 of 9 Table 1                                          l CONTRIBUTION TO THE LOCA DOSE EXPOSURES FOR A MAXIMUM MSIV                      j LEAK RATE OF 100 sCfh PER STEAM LINE, 200 sCfh TOTAL DUANE ARNOLD ENERGY CENTER Whole Body      Thyroid     Beta (rem)      (rem)     (rem)
Exclusion Area         A)  10 CFR 100 Limit                  25          300
* Boundary (2-Hour)
B)  Previous Calculated Doses **      7.64        28.6 C)  Previous Calculated Doses w/o     7.64        28.6   ****
MSIV Leakage D)  Contribution From MSIVs at      <0.01        0.05 100 scfh, 200 scfh total E)  New Calculated Doses              7.65       28.7
. I.ow Population Zone   A)  10 CFR 100 Limit                  25          300      *
(30-Day)
B)  Previous Calculated Doses **      2.86       24.4 C)  Previous Calculated Doses w/o     2.86       '24.4    ****
MSIV Leakage D)  Contribution From MSIVs at        0.04        18.03 100 scfh, 200 scfh total E)  New Calculated Doses              2.90       42.5 Control Room (30-      A)  GDC-19                              5          30   30/75***
Day)
B)  Previous Calculated Doses **      0.62       0.20     8.6 C)  Previous Calculated Doses w/o     0.62       0.20    8.6****
MSIV Leakage D)  Contribution From MSIVs at        0.03      22.39     0.37 100 scfh, 200 sefh total E)  New calculated Doses              0.65        22.6    9.0 Technical Support      A)  GDC-19                            5           30    30/75***
Center (30-Day)
Previous Calculated Doses
* 4.76      *****     *****
B)
C)  Previous Calculated Doses w/o  4.76****
MSIV Leakage D)  Contribution From MSIVs at        0.02       2.15     0.35 100 scfh, 200 scfh total New Calculated Doses              4.78      *****     *****
E) 0*        No limit specified.
LPZ and EAB from APED-D21-029 (1658 mwt uprate) April 1984; Control Room doses from UFSAR Table 6.4-1.
o**   75 if prior commitment has been made to use protective clothing.
    ****  Contribution of MSIV leakage to doses are unknown; and, therefore, are conservatively assumed to be zero.
    ***** Not Calculated.
J

RTS-232                                                                                                                                                                   Attachment 2 to NG-94-2629
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                                                                                        > OFFGAS SYSTEM NOTES:
                              = ALTERNATE DRAIN PATH                                                (BECH-M141)         1. V03-0004 AND V03-0005          4. MAIN STEAM LINES BETWEEN WILL BE CONVERTED TO               THE MSIV UP TO THE STOP
                              = PRIMARY DRAIN PATH                                            STEAM JET AIR               MOTOR OPERATED VALVES.             VALVE. INCLUDING BRANCH
             ~                                                                          = EJECTOR IE008B 2. ESSENTIAL POWER WILL BE                          LINES 3' AND LARGER HAVE
_A = SEISMIC ADEDUACY                                                                                                                BEEN SEISMICALLY ANALYZED.
WALKDOWN BOUNDARY                                                 (BECH-M103             PROVIDED TO THE FOLLOWING
[          $                                                                                                            VALVES: V03-0004. V03-0005     5. ALL CONNECTIING BRANCH (f)                                                                              STEAM JET AIR               MO-1169. M0-1170. MO-1043.         LINES SUCH AS HPCI. RCIC.
  .:z:
* EJECTOR.1E008A                  Mo-1044                            EcuALIz1NG HEADER ORAIN, t-r3                        PRIMARY                                                              (BECH-M105)          3.LINE UP DEPICTS A POST             (2' AND UNDER). VENTS AND
:r=.-      C                CONTAINMENT cEE
                                                  "   NOTES                           EE NOTES                            LOCA LINE UP. OPERATOR             INSTRUMENTATION TAPS ARE
:::o       q                                        1&2       w-                       1&2                              ACTION IS REQUIRED TO              NOT DEPICTED BUT WERE INCLUDED IN THE EOE
:2":
c3         H                                                                 V V03-0005                                 POSITION THE FOLLCWING SEISMIC ADE00ACY f                                     V03-0004{                   g                                          vtv S: v03-0004.V33-0005.
{          *
(NO)                              (NO)
IC213B               Mo-1169. Mo-1043. Mo-1044 Cv-1064 EVALUATION.
                                                               ?              ?           TURBINE BUILDING                                               6 NO: NORMALLY OPEN
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RTS-232                                              Attachment 3 to NG-94-2629 Page 1 of 1 PROPOSED CHANGE (RTS-2321 TO THE DUANE ARNOLD ENERGY CENTER TECHNICAL SPECIFICATIONS The holders of license DPR-49 for the Duane Arnold Energy Center propose to amend Appendix A (Technical Specifications) to said license by revising the attached pages as indicated. The List of Affected Pages is given below.
LIST OF AFFECTED PAGES 111*
3.7-5*
3.7-19a*
3.7-49a*
* Affected by RTS-246, submitted March 27, 1992 (NG-92-1238)

==SUMMARY==
OF CHANGES:
The following list of proposed changes is in the order that the changes appear in the Technical Specifications (TS).
Page                         Description of Changes ill                 This page of the Table of Contents is being revised to reflect the deletion of TS Sections 3.7.E and 4.7.E. Section 3.7.F is also being listed to correct its omission from a previous amendment.
3.7-5               This proposed change increases the allowable leak rate specified in TS 4.7.A.2.c.3 from 11.5 standard cubic feet per hour (scfh) for any one main steam isolation valve (MSIV) when tested at 24 psig to 100 scfh for any one MSIV with a total maximum pathway leakage  1 rate of 200 scfh through all four main steam   :
lines when tested at 24 psig. If an MSIV exceeds 100 scfh, it will be restored to 5 11.5 scfh.
l 3.7-19a, 3.7-49a    This proposed change deletes TS 3.7.E and      l 4.7.E and the related section 3.7.E and 4.7.E of the Bases. It involves eliminating the MSIV leakage control system (LCS) requirements from the TS.                      I
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