HL-4468, Forwards Response to NRC 931229 RAI Re 931001 Request to Revise TS for Increase in Allowable MSIV Leakage Rate & Deleton of MSIV Leakage Control Sys.Revised Proposed TS Pages & Suppl Piping Earthquake...Data Also Encl
| ML20059E919 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 01/06/1994 |
| From: | Beckham J GEORGIA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20059E921 | List: |
| References | |
| REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR HL-4468, TAC-M87850, NUDOCS 9401130053 | |
| Download: ML20059E919 (27) | |
Text
__.
Goorgia Power Company 40 invemess Center Parkway g Post Ofhce Box 1295 O Birmingham, Alabarna 35201 Tdphone 205 877-7279
&A J. T, Beckham, Jr. (jC()[yill POWCf Vice President Nucioar Hatch Propict fN < W ' " t 9 da : > :
January 6, 1994 Docket No. 50-366 IIL-4468 TAC No. M87850 U S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D C. 20555 Edwin 1. llatch Nuclear Plant - Unit 2 Request to Revise Technical Specifications:
Increase in Allowable MSIV Leakage Rate and Deletion of the MSIV In: age Control System Gentlemen:
By letter dated October 1,1993, Georgia Power Company (GPC) submitted proposed changes to the Plant llatch Unit 2 Technical Specifications, Appendix A to Operating License NPF-5, to increase the allowable main steam isolation valve (MSIV) leakage and delete the requirements for the currently installed MSIV leakage control system, On December 10, 1993, GPC representatives and consultants met with the Nuclear Reactor Regulation (NRR) stafTto discuss the proposed changes and to provide responses i to the NRR staffs questions and concerns. During the meeting, the NRR stafTexpressed comments relative to the radiological dose assessment corresponding to a total leakage of 400 scfh, a postulated failure of normally closed valve 2B21-F021 in the main steam line drain header downstream of the MSIVs, the small diameter piping interconnected with the !
main condenser, and the piping suppon margin assessment. Additionally, the NRR statT requested a clarification relative to the criteria for evaluating the anchorage for equipment and piping. By letter dated December 29,1993, the NRR stalTrequested GPC to provide a response to the above comments.
As discussed in the December 10,1993 meeting, GPC is expediting the approval of the proposed Technical Specifications changes by revising the October 1,1993 submittal, utilizing the more conservative assumptions used in the NRR stalTs analysis, thereby ,
providing further assurance offsite doses will not exceed 10 CFR 100 hmits. !
l 1
I 9401130053 940106 f.
l
9 Georgia Power d I
U.S. Nuclear Regulatory Commission Page 2 January 6, 1994 l
Consequently, the proposed increase in MSIV leakage was revised to 100 sc01 per MSIV l with the total leakage not to exceed 250 scfh instead of 400 scfh as originally proposed. l Also, GPC has determined that there is reasonable assurance valve 2B21-F021 will )
perform its intended function to establish a drain path to the condenser. As discussed l during the meeting, GPC identi6ed a second alternate drain path from the main steam i lines to the main condenser to resolve the issue associated with a postulated failure of j valve 2B21-F021. 1 As discussed during the meeting, GPC will submit a justification for more realistic i atmospheric dispersion factors. This justification, which will be provided in the Unit I submittal to relax MSIV leakage, will also address Unit 2.
I During the December 10, 1993 meeting, the NRR stafT requested GPC to clarify the ;
method used to evaluate concrete anchor bolts. In response, the evaluation of concrete anchor bolt capacity for piping supports and equipment used the criteria contained in the i Seismic Qualification Utility Group Generic Implementation Procedure for resolution of .
Unresolved Safety Issue A-46.
1 Enclosure 1 provides detailed descriptions of GPC's response to the requested additional j information. Enclosure 2 contains the revised basis for change request for proposed Change 1 and proposed Change 2. Enclosure 3 contains the revised significant hazards considerations for proposed Change I and proposed Change 2 that supersede the corresponding changes provided in GPC's submittal dated October 1,1993. Proposed Change 3 and Proposed Change 4 contained in the October 1,1993 submittal do not 1 require any revision. The conclusions of the significant hazards evaluations contained in the October 1,1993 submittal are unalrected by the revised proposed Technical Specifications changes provided in this submittal. Enclosure 4 provides page change instructions for incorporating the proposed changes. The revised, proposed Technical Specifications pages, along with a marked-up copy of the current Technical Specifications pages, follow Enclosure 4.
5 U.S. Nuclear Regulatory Commission Page 3 Janvary 6, 1994 i
c i hir. J. T. Beckham, Jr. states h: is duly authorized to execute this oath on behalf of Georgia Power Company, and to the best of his knowledge and belief, the facts set fonh in this letter are true.
GEORGIA POWER COh1PANY By: h _
J. T. Beckham, JF i Sworn to and subscribed before me this 6 dayo Mm/ ,1994 W h. k W _
Notary Public My Commission Expires /cg. 8,1905 004468 *
Enclosures:
- 1. Response to Request for Additional Information !
- 2. Basis for Change Request
- 3. 10 CFR 50.92 Evaluation
- 4. Page Change Instructions :
1 cc: Georgia Power Company H. L. Sumner, Jr., Nuclear Plant General hianager l NORh1S U.S. Nuclear Rerulatory Commission. Washington. D.C.
K. N. Jabbour, Licensing Project hianager - IIatch U.S. Nuclear Regulatory Commission. Recion II-S. D. Ebneter, Regional Administrator :
L. D. Wert, Senior Resident Inspector - IIatch State of Georgia hir. J. D. Tanner, Commissioner - Department of Natural Resources
4 g-W+-- 0 16 __4Jul.J4 u __e 44 quu- - ., .4_s, aA 4 - _-- ..--. g ,
s 1
Enclosure 1 Edwin I. Ilatch Nuclear Plant Request to Revise Technical Specifications:
Increase in Allowable MSIV Leakage Rate and Deletion of the MSlV Leakage Control System I
Response to Request for Additional Information l I
l On December 10,1993, Georgia Power Company (GPC) representatives and consultants l met with the Nuclear Reactor Regulation (NRR) staff to discuss GPC's request to revise l the Unit 2 Technical Specifications to delete the main steam isolation valv.e (MSIV) leakage control system and to increase the allowable MSIV leakage rate. During the meeting, the NRR. staff requested additional information related to the radiological dose calculations, a postulated failure of normally closed valve 2B21-F021 in the main steam line drain header downstream of the MSIVs, and the functional requirements of small '
diameter piping interconnected with the main condenser. GPC's response is as follows:
- 1. NRR StafrComment:
The licensee should address the methodology used to determine X/Q.
GPC Response:
l The recalculated loss of coolant accident (LOCA) doses, excluding MSIV leakage and the contribution to the LOCA dose exposures for a maximum MSIV leak rate of 100 scfh, were provided in Tables 1 and 2, respectively, of GPC's submittal dated October 1,1993. The doses were conservatively calculated by a methodology for ;
using the isolated condenser method for MSIV leakage treatment developed by General Electric and used the X/Q values described in section 2.3 of the Unit 2 Final Safety Analysis Report (FSAR). These calculations concluded that the proposed total j leakage of 400 scfh resulted in LOCA doses that remained within the guidelines of ;
10 CFR 100 for offsite doses and 10 CFR 50, Appendix A, for the control room and
- technical support center doses. During the meeting, the NRR stafT presented the results of their independent radiological dose calculation corresponding to an MSIV leakage rate of 400 scfh. The staffs calculation used the X/Q values contained in the
- Safety Evaluation Report (SER) for operation of Unit 2. These values differ- j significantly from those in FSAR section 2.3. As a result, the NRR staffs calculation predicted ofTsite doses in excess of the guidelines of 10 CFR 100. The i E
l 4468 El-1
Enclosure 1 Response to Request for AdditionalInformation i NRR staff stated that the X/Q values contained in the SER were the only values available for their use and requested GPC to either submit new X/Q valves, calculated i consistent with the guidelines of Regulatory Guide 1.145 or to consider lower MSIV leakage rates.
Georgia Power Company has determined that a revision to the proposed total MSIV leakage is the appropriate resolution to support an expedient review by the NRR staff. j Consequently, Enclosure 2 contains the revision to the proposed changes to the .
Technical Specifications for an allowed leakage of 100 scfh per MSIV with a total-allowed leakage not to exceed 250 scfh GPC anticipates that the referenced X/Q values will be recalculated and submitted in the future fbt NRR stafTreview as part of l the proposed Technical Specifications change to increase the allowable MSIV leakage for Unit 1.
i
- 2. NRR StafTCommsnt:
The licensee should address the single failure criterion with regard to the drain path valves. l GPC Response:
GPC proposes to delete the leakage control system requirements from the Technical Specifications and to use the main steam drain lines and the isolated main condenser as an ahernate treatment pathway for MSIV leakage. The primary drain path employs a main steam line drain downstream of the MSIVs. There are two motor operated valves in this drain line. One valve,2B21-F020 is normally open and will fail in the "as is" position upon loss of power. Therefore, it is very unlikely valve 2B21-F020 would ever be in the closed position if the main steam lines were required for MSIV_ leakage treatment. The other valve,2B21-F021, is normally closed and is required to open to ;
use this drain header. During the December 10,'1993 meeting, the NRR staff l expressed a concern that a postulated failure of valve 2821-F021 to open would prevent the establishment of a drain path to the condenser.
In response, GPC has concluded that there is reasonable assurance that the subject valve will perform its intended function to establish the drain path. This valve can be powered from either division of emergency AC power and can be opened during a loss of offsite power. Also, the maintenance history for approximately the last 10 years-4468 El-2
Enclosure 1 Response to Request for Additional Information was reviewed to evaluate the valve's performance. The review concluded that no failure that would have prevented the establishment of this drain path was identified.
To provide additional assurance, GPC will include valve 2B21-F021 in the inservice testing program to perform a stroking surveillance on a quarterly basis.
Additionally, GPC has verified that a second drain pathway is available to convey MSIV leakage to the isolated cc,ndenser if valve 2B21-F021 fails to open. This drain path is located downstream of the primary drain path and originates frorn the main steam line drain pots. This second drain path was previously included in the seismic verification scope. The alternate path has a large 0.8-inch restricting orifice in a bypass line around a normally closed valve in the drain line. Consequently, if valve 2B21-F021 failed to open as required, the second drain path, along with the 0.105-inch restricting orifice in the bypass line around valve 2B21-F021, would be available to convey MSIV leakage to the isolated condenser. This second path (consisting of the two orifices) will convey essentially all of the MSIV leakage to the condenser.
Consequently, the radiological dose assessment for this alternate pathway is essentially equivalent to the dose assessment for the primary path. -
- 3. NRR StafTComment:
The licensee should provide a more complete database for small bore piping. [
t GPC Response:
Subsequent to the December 10,1993 meeting, the NRR staff requested additional information relative to the functional requirements for mechanisms to transport the MSIV leakage to the isolated condenser and the performance requirements for the -
small diameter interconnected piping. The NRR stafT also requested additional data from the earthquake experience data base on small diameter piping. GPC's response is as follows:
The alternate method for MSIV leakage treatment uses the main steam drain lines to convey the MSIV leakage to the isolated main condenser. This leakage treatment method takes advantage of the large volume in the main condenser to provide hold-up and plate-out of the fission products that may leak from closed MSIVs and provides efTective fission product attenuation in the condenser such that the consequences of MSIV leakage can be significantly reduced. MSIV leakage that enters the condenser 4468 El-3 ,
w s-~ , , -
a m.,,a- -- ,- --, , e . ,-- a - - - , ..,n .. ,- . _ .------- . - - - - -
r l
Enclosure 1 Response to Request for AdditionalInformation is ultimately released to the turbine building through the low pressure turbine seal after significant plate-out ofiodine.
The main steam drain lines direct the leakage to the condenser. The post design basis accident containment pressure, along with gravity flow, provides the transport mechanisms. Neither condenser vacuum nor auxiliary equipment, such as the steam jet air ejectors, is needed to convey the leakage to the condenser.
The small diameter piping included in the main steam drain path, interconnect'ed with the main steam drain path, or interconnected with the main condenser is required to retain its integrity. Consequently, this piping is included in the seismic verification scope to ensure that the piping will establish a path or boundary to convey the leakage to the condenser.
The seismic verification scope for piping consists of the following:
- 1. The main steam drain path to the condenser for any leakage past the isolated outboard MSIVs.
- 2. Main steam piping from the outboard MSIVs to the main steam stop valves.
- 3. Main steam bypass piping from the main steam lines to the bypass valve chest.
- 4. Additional interconnected piping within the Seismic Verification Boundary, such as steam jet air ejector lines, steam supply to the reactor feed pump turbine, and various drains.
The " additional" interconnected piping is included to ensure that boundaries are established within the drain path.
In response to the request for additional data, supplemental piping earthquake performance data from the earthquake experience data base is provided as an attachment to this enclosure.
9 4468 El-4
Enclosure 1 Response to Request for Additional Information
- 4. NRR StatTCommeE:
In view of the support damage experienced by some seismic database plants, the licensee should demonstrate that similar damage to the supports is not likely to occur at llatch Unit 2 as a result of the design basis earthquake. Furthermore, the licensee should expand the discussion of the suppon margin assessment presented in sections 4.1.1.4 and 4.2.1.4 of the October I,1993, submittal to include: (1) a justification for the use of only a vertical floor response spectrum of 0.75g, and (2) a justification for the range of 0.75 g to 1.5 g for the high confidence oflow probability of failure of the supports.
GPC Respon_g:
This was explicitly done at llatch Unit 2 by performing the following steps:
- a. A comparison was made of the liatch Unit 2 plant features to NEDC-21858P Appendix D, " Performance of Condensers and Main Steam Piping in Past Earthquakes," to verify representation by the earthquake experience data that demonstrated good seismic performance.
- b. A review of the design basis of the llatch plant features associated with the MSIV leakage closure issue was made to insure that commercial codes, standards, and practices were utilized in the designs that have demonstrated good historical seismic performance.
- c. Most importantly, a seismic walkdown was performed of all systems and components associated with the MSIV leakage closure issue. The purpose of the walkdown was to: (1. physically verify that the liatch plant features have the attributes similar to those in the earthquake experience data base that . have demonstrated good seismic performance; and (2. look for any seismic vulnerabilities that could potentially bring into question the seismic ruggedness of a system or component. The support damage experienced by some seismic database plants is used to help identify potential seismic vulnerabilities during the walkdown. Potential vulnerabilities were identified as " outliers"' during the walkdown for resolution later.
- d. The " outliers" were evaluated and modifications were developed as necessary to ensure good seismic performance consistent with earthquake experience.
4468 El-5
Enclosure 1 Response to Request for Additional Information Section 4.1.1.4 states that the main steam and the turbine bypass piping as defined is analyzed, designed, and constructed to seismic Category I criteria and meets the requirements of the ASME Code, Section 111, Class 2. Based on seismic margins research, piping systems designed to these standards have very large seismic margins.
Section 4.2.1.4 discusses the seismic margin assessment of a representative and bounding ,
suppon for the main steam drain line The purpose is to provide additional assurance of good seismic performance beyond the review of its design basis and the results of the seismic walkdown. The following is additional discussion of the seismic margin assessment performed:
- a. Only the peak spectral acceleration of the vertical floor response spectrum, i.e., 0.75g, was used since the supporting system for the main drain line is dead load hangers whose capacity would be challenged only by the vertical earthquake component of motion. The support system for the main drain line; i.e., dead load hangers, is typical of those in the earthquake experience data base that have performed well in an earthquake. The purpose of the seismic margin assessment is to show there is significant margin on the capacity of the dead load supports such that one has additional assurance the piping will retain its supporting system during an earthquake, i i.e., position retention of the piping is maintained. {
l
- b. The range of the high confidence oflow probability of failure (IICLPF) of the supports ')
is based on simply following the Conservative Deterministic Failure Margin Approach l (CDFM) as specified in EPRI NP-6041 report. A conservative estimate of the suppon test load capacity of the standard dead load vertical support is determined and the i carthquake level is determined that would produce a demand equal to the support .
capacity. The carthquake level, in terms of the peak ground acceleration level, is l defined as the HCLPF value. The HCLPF values were found to be significantly higher l than the Hatch design basis earthquake (DBE) level, therefore, providing additional assurance of good seismic performance at the IIatch DBE level.
- 5. NRR StafrCommen_t:
The licensee should document that anchorages for piping and equipment conform to :
the criteria found in Section 4, " Screening Verification and Walkdown," of the SQUG !
Generic Implemutation Procedure for Seismic Verification of Nuclear Plant Equipment.
4468 El-6
i i
l
. . 1 l
l Enclosure 1 Response to Request for Additional 1nformation GPC Response:
The evaluation of concrete anchor bolt capacity for piping supports and equipment used the criteria contained in the Seismic Qualification Utility Group Generic Implementation Procedure for Resolution of Unresolved Safety issue A-46.
b I
f t
l 4468 El-7 i
E Attachment to Enclosure 1 l
Supplemental Piping Earthquake Performance Data I
i e
P i
l I
I 4468 El-8
Enclosure 2 Edwin I. Hatch Nuclear Plant - Unit 2 Request to Revise Technical Specifications: i Increase in Allowable MSIV Leakage Rate and l Deletion of the MSIV Leakage Control System Basis for Change Reqqcst Proposed Change 1 This proposed change revises Technical Specification 3.6.1.2.c to increase the allowable main steam isolation valve (MSIV) leakage from "11.5 scf per hour for any one main steam isolation valve when tested at 28.8 psig" to "when tested at 28.8 psig,100 scf per hour for any one main steam isolation valve and a combined maximum pathway leakage rate of 250 scf per hour for all four main steam lines." The proposed change also changes the associated Action for Technical Specification 3.6.1.2 from "11.5 scf per hour for any one MSIV" to "100 scf per hour for any one MSIV or a total maximum pathway leakage rate of > 250 scf per hour for all four main steam lines." If the leakage for any MSIV exceeds 100 standard cubic feet per hour (scfh), it will be restored to 115 scfh. If the total maximum pathway MSIV leakage for all four main steam lines exceeds 250 scfh, the necessary MSIVs will be restored such that the maximum pathway leakage is no more than 250 scfh.
Basis for Proposed Change 1 The current Technical Specifications allowable MSIV leakage rate is extremely limiting and routinely requires repair and retest of the MSlVs. This significantly impacts the maintenance work load during plant outages and contributes to outage extensions. The outage planning group at Plant Hatch typically schedules several days of contingency for repair and retest of the MSIVs. The proposed increase in the allowable MSIV leakage would reduce the need for repair and, thereby, reduce dose exposures to maintenance personnel consistent with As Low As Reasonably Achievable principles. Finally, there have been many Licensee Event Reports (LERs) written within the industry for MSIV leakage which fails to meet the current Technical Specifications limit including a number for Plant Hatch. The generation of such LERs represents .a needless expenditure of resources which could be better spent on issues of greater safety significance.
Failures of MSIVs to meet the current Technical Specifications 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.
4468 E2-1
Enclosure 2 Basis for Change Request The BWROG, with Georgia Power Company (GPC) participation, studied the issues regarding A1SIV leakage rates, their causes, and available alternatives. The results of the llWROG study are provided in NEDC-31858P, "BWROG Report for increasing N1SIV Leakage Rate Limits and Elimination of Leakage Control Systems," Revision I and are also summarized in NUREG-1169. In response to Generic Issue C-8, "h1SIV Leakage and LCS Failure," the BWROG has recommended corrective actions and maintenance practices to reduce the h1SIV leakage rates.
A survey conducted by the BWROG of h1SIV LLRT results between 1984 and 1988 indicated the implementation of industry and BWROG actions has been effective in ,
reducing the NISIV leakage rates. Of key importance was the reduction in the number of valves which experienced substantially high leakage rates. liowever, the survey also concluded about 23 percent of the total "as found" h1SIV leakage rates still exceeded the limit of 11.5 scfh and about 10 percent exceeded 100 scfh. The h1SlV leakage performance at Plant Ilatch is representative of the generic htSIV leakage data collected by the NRC and the BWROG.
Despite the improvement in leakage performance, h1SIV leakage rates still frequently :
exceed the current Technical Specifications limit and the resultant maintenance problems, l although less severe, remain as a significant issue. Furthermore, based on extensive evaluation of valve leakage data, the BWROG has found disassembling and refurbishing j the h1SIVs to meet very low leakage limits frequently contributes to repeating failures. In most cases, machining of the valve seat is required to reduce the leakage to an acceptable level. 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 having to disassemble the valves and refurbish them for minor leakage, the utility may avoid introducing one of the root causes of recurring valve leakage problems which lead to later LLRT failures and the possibility of compromising plant safety.
The current Technical Specifications allowable h1SIV leakage rate (11.5 scfh) is excessively conservative considering the valve's physical size and operating characteristics (large size and fast-acting), Additionally, the existing turbine building equipment was not considered at the time the leakage limit was established. Based on the in-depth evaluation of h1SIV leakages, the BWROG has concluded leakage rates up to 500 scfh are not indicative of substantial mechanical defects in the valves which would challenge the l
1 4468 E2-2 i
. . _ _ , , . . _ . . , - I
i Enclosure 2 '
Basis for Change Request l
capability of the valves to fulfill their safety functions of isolating the steam lines. ]
Furthermore, valve manufacturers have stated leakage rates up to .200 scfh can occur without having a major valve defect. Therefore, the proposed increase from 11.5 scfh per i MSIV to 100 scfh per MSIV with a total maximum pathway leakage of 250 scfh for all four main steam lines will not alTect 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. Based on the BWROG study, the proposed increase in the allowable leak. age rate will increase the chance for successful LLRT results
{
to greater than 90 percent, up from the 77 percent success rate at the current limit of ;
11.5 scfh. At Plant Hatch, the increase in successful local leak rate testing will ;
significantly reduce MSIV maintenance cost, reduce dose exposure to maintenance personnel, reduce outage durations, extend the effective service life of the MSIVs, and ,
minimize the potential for outage extensions. j New control room, technical support center (TSC) and ofTsite doses have recently been recalculated for a postulated loss of coolant accident (LOCA), as described under the justification for Change 2 on pages E2-6 and E2-7. The new doses are presented in l' Table 1 of this enclosure. The radiological dose methodology developed by General Electric (GE) for the BWROG was used to calculate the effects of the proposed MSiv l leakage rate. The revised LOCA doses are the sum of the recalculated doses and the newly calculated MSIV leakage doses as shown in Table 2 of this enclosure. These i analyses demonstrate that the proposed MSIV leakage rate results in acceptable dose exposures for the control room, TSC, and ofTsite boundaries.
Proposed Change 2 ;
This proposed change deletes Technical Specification 3/4.6.1.4 and Bases section ,
3/4.6.1.4 and permits climination of the MSIV leakage control system (LCS). A more reliable alternative treatment method of MSIV leakage is proposed.
4468 E2-3 4
y - ..e,. -
-.n . , . , , , , . .. -- _----- - --
i
?
Enclosure 2 Basis for Change Request b
Basis for Proposed Change _2 L s
As a condition for obtaining an operating license for Plant Hatch Unit 2, a safety-related LCS was required to be installed which satisfied the guidance of Regulatory Guide 1.96,
" Design of Main Steam Isolation Va!ve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants," to reduce the potential radiological consequences of post accident MSIV leakage.
In 1983, Generic Issue C-8 was established to track the resolution of an NRC concem 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 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 Plant Hatch 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 as described in section 15.1.39 of the Final Safety Analysis Report (FSAR).
Georgia Power Company proposes to delete the LCS requirements from the Technical Specifications and to use the main steam drain lines and the isolated main condenser as an alternate method for MSIV leakage treatment. The BWROG, with GPC participation, has evaluated several alternate MSIV leakage treatment methods and has recommended the isolated condenser for MSIV leakage treatment. This leakage treatment method takes j advantage of the large volume in the isolated main condenser to hold up the release of any '
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 l capability at high MSIV leakage rates and will assure a reliable and efTective method is j available for treating any potential MSIV leakage during a postulated LOCA. GPC will I also incorporate the applicable alternate leakage treatment methods into the Operating and/or Emergency Operating Procedures as appropriate.
Two motor-operated valves are located in the main steam line drain header downstream of the MSIVs. One of the valves is normally closed. However, it is powered from an emergency AC power bus and can be opened during a loss of offsite power. The other valve is a normally open valve which is powered from the normal power system. Since this valve is normally open and will fail "as is" on loss of power, it is very unlikely it would ever be in the closed position if the main steam line drains were required for treatment of MSIV leakage.
4468 E2-4 .
1
.w , - r --
m -ae v--- -
s- -.m.-
-rs e,e . w rw
r i
i Enclosure 2 Basis for Change Request t
Georgia Power Company has verified a second drain pathway is available to convey MSIV !
leakage to the isolated condenser if valve 2B21-F021 fails to open. This drain path is :
located downstream of the primary drain path and originates from the main steam line j drain pots. This second drain path was previously included in the seismic verification scope. The alternate path has a large 0.8-inch restricting orifice in a bypass line around a i normally closed valve in the drain line. Consequently, if valve 2B21-F021 failed to open as required, the second drain path, along with the 0.105-inch restricting orifice in the i bypass line around valve 2B21-F021, would be available to convey MSIV leakage to the ,
isolated condenser. The second path (consisting of the two orifices) will convey essentially all of the MSIV leakage to the condenser. Consequently, the radiological dose assessment for this alternate pathway is essentially equivalent to the dose assessment for the primary path.
i in addition to resolving the concern identified in Generic Issue C-8, the proposed deletion j of the LCS requirements from the Technical Specifications 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 l preventive maintenance. The system has extensive logic and instrumentation which requires frequent calibration to meet the Technical Specifications requirements. The BWROG evaluated recent LCS performance data; the results are shown in NEDC-31858P. The evaluation indicates the LCS is extremely diflicult to maintain, and .
as a resuh of maintenance requirements, plant shutdowns and startup delays have occurred I within the industry.
i While the LCS has not caused any plant shutdowns at Plant Hatch, it has caused the plant to be placed in a limiting condition for operation (LCO) on a number of occasions.
Twenty-nine component failures in the LCS, which required reporting under the Nuclear Plant Reliability Data System, have occurred. Many of the failums required extensive component out of service times, and significant cost in the form of resources and personnel exposure was incurred.
An event involving the failure of an LCS blower in 1989 is an example of the high cost of maintaining the system. The blower failure required entry into a 30-day LCO. A replacement assembly was not available at the plant site or within the GPC system. The motor / blower assembly was manufactured by Siemens for GE; however, Siemens discontinued the. manufacturing of the motor / blower in the mid 1970s. A search throughout the industry for a replacement unit was initiated concurrently with efforts to repair the existing unit. GE was able to locate and dedicate the necessary parts to 1 i
4468 E2-5
Enclosure 2 Basis for Change Request ;
assemble a qualified replacement unit. The cost of the replacement unit delivered on site was approximately $270,000. A similar failure in the future could create an even worse condition since GE has indicted that no more units are available. Only a few spare units exist, and utilities who own them are unwilling to release : hem to other BWR owners.
The BWROG has evaluated the availability of the main steam system piping and conde user !
alternate treatment pathway for processing MSIV leakage. It was determined the probability of a near coincident LOCA and seismic event is much smaller than other plant i safety risks. The BWROG has also determined that main steam piping and condenser designs are extremely rugged, and the B31.1 design requirements typically used for nuclear plant system designs contain a good deal of margin. The Plant llatch Unit 2 main steam lines are seismically qualified up to the turbine stop valves.
To furtherjustify the capability of the mam steam pipmg and condenser alternate treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non-seismically designed piping and condensers (in past earthquakes).
That study summarized the data on the performance of main stem piping and condensers in non-nuclear applications which experienced strong motion earthquakes and compared those piping and condenser systems with the piping and condenser systems typically used in GE BWRs in the United States. The results of the comparison strengthen the position l that main steam piping and condensers employed in GE BWRs would maintain their pressure retention function during a design basis earthquake. It is, therefore, concluded ,
the possibility of a failure which could cause a loss of steam or condensate in the Plant llatch main steam piping or condensers in the event of a design basis earthquake is extremely low, and such a failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented. This conclusion is consistent with l NUREG/CR 4407, " Pipe Break Frequency Estimation for Nuclear Power Plants," dated i
May,1987, which reported no observed failures in main steam piping over 313 reactor years of operation. Therefore, the isolated condenser alternate MSIV leakage treatment path at Plant Hatch is considered appropriate for the minimization of potential radiological consequences of a design basis LOCA.
i 1
As additional verification of the scismic adequacy of the proposed alternate MSIV leakage ;
treatment system, the main steam lines downstream of the outboard MSIVs, main steam i drain lines, main steam branch lines, and condenser at Plant Hatch Unit 2 have been i
walked down by seismic review teams composed of members with extensive seismic I evaluation and analyses experience. Five of the six individuals who participated in the walkdown had, at the time of the walkdown, received the Seismic Qualification Utility Group (SQUG) training as specified in the Generic implementation Procedure (GIP).
Conditions considered during the walkdown include failure and proximity impact, 4468 E2-6
- ..- ~ - . .- .- - . . - - . . . .-. .
, u.
Enclosure 2 Dasis for Change Request l
ditTerential seismic anchor motions, and equipment anchorage. The potential outliers identified during the walkdown have been resolved or will be resolved prior to implementation of the modification to remove the LCS. A report of the equipment walkdown, including a description of the outliers and their resolution, was provided in the October 1,1993 submittal. In addition, the condenser installed at Plant Hatch was confirmed to fall within the bounds of design characteristics found in selected conventional power plant condensers included in the carthquake experience data base of Appendix D to NEDC-31858P. The main steam line piping including the piping between the outboard A1SIVs and the turbine stop valves is seismically qualified for a design basis earthquake. j Therefore, the main steam lines, the condenser, and the main steam drain lines proposed to be used in the alternate h1SIV leakage treatment process at Plant Hatch are believed to be seismically rugged and capable of performing the proposed passive function. Based on the -
above information, it is concluded that the main steam lines, main steam drain lines, and the condenser meet the general intent of Appendix A to 10 CFR 100 to perform their safety function during and following a potential design basis earthquake.
Standard conservative assumptions were used to calculate ofTsite, control room, and TSC doses, including the doses due to hiSIV leakage, which could potentially result from a postulated design basis LOCA at Plant Hatch. The results of those calculations are currently described in section 15.1.39 of the Hatch Unit 2 FSAR. The calculated control room, TSC, and offsite doses due to a LOCA are shown in FSAR Tables 15.1-28, 15.1-28a, and 15.1-36, respectively. The control room, TSC, and ofTsite doses resuhing from a postulated LOCA have recently been recalculated using currently accepted iodine dose conversion factors. The control room and TSC doses were calculated using the guidance in Regulatory Guide 1.109, " Calculation of Annual Doses to hian from Routine Releases of Reactor Efiluents for the Purpose of Evaluating Compliance With 10 CFR-Part 10, Appendix 1." The otTsite doses were calculated using the guidance contained in ,
EPA Federal Guidance Report No. I1, " EPA-520/1-88-020 Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, i Submersion, and Ingestion." The results of the recalculations are presented in Table 1 of this submittal. An FSAR revision to incorporate the newly calculated values is being prepared and will be issued following NRC approval of this license amendment request.
The radiological dose methodology developed by GE for the BWROG is documented in Appendix C of NEDC-31858P. This radiological analysis was used to calculate the effects of the proposed allowable hiSIV leakage rate in terms of control room, TSC, and ofTsite doses. The revised LOCA doses are the sum of the recalculated LOCA doses as shown in Table 1 and the newly calculated doses due to the increased allowable h1SIV leakage. The ]
TSC doses due to h1SIV leakage are considered to be very conservative. l 4468 E2-7 l
l
i l
j Enclosure 2 Basis for Change Request It is not expected there will be any radioactive releases to the TSC due to MSIV leakage during the first 30 minutes following a LOCA event, since it would take considerable time for the MSIV leakage to travel through the main steam lines and main steam line drain system to the condenser, into the turbine building, and fmally to the atmosphere and TSC.
Ilowever, it was conservatively assumed the 30-day integrated dose of 3.01 rem due to a total MSIV leakage of 250 scfh from all four main steam lines could be received by personnel who start the filter system and immediately enter the TSC any time afler the LOCA occurs. The dose calculations were made using control room occupancy factors specified in SRP 6.4.
Table 2 shows the calculated dose exposures from the BWROG radiological analysis for Plant Hatch. Regulatory limits and calculated doses from LOCA radiological analysis are also included in Table 2 for comparison purposes. This analysis demonstrates that a total leakage rate of 250 scfh (with the deletion of the existing 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. .
I)eletion of the LCS will reduce the overall dose rates and eliminate the system's impact on ,
refueling and maintenance outage activities at Plant Hatch. The proposed ahernate method (main steam lines and condenser) for MSIV leakage treatment will also eliminate the concern regarding LCS effectiveness at higher MSIV leakage rates.
I 4
9 d
4468 E2-8
l l
TABI,E1 l
Recalculated Loss-of-Coolant Accident Doses (rem) Excluding 51SIV Leakage IIntch Nuclear Plant Control Room Thyroid 10.0 Whole Body 0.066 Technical Support Center Filter Start Time
- Thyroid Whole Body 0 min 2.96 0.0328 10 min 21.06 0.0316 30 min 17.46 0.0286 In all cases, doses are based on no personnel entry into the TSC until after tilter initiation unless protective respirators are worn.
OfTsite (2 hr) EAR (2 hr)EAB (30 days)LPZ (30 days)LPZ Pathway _ _ Thyroid Whole Body Thyroid Whole Body Drawdown 64.30 0.970 64.3 0.97 .
(120 s)
Containment 0.77 0.053 5.45 0.11 ,
Leakage (1.2 %)
Bypass (0.9% 25.40 0.200 100.70 0.36 of 1.2 %)
TOTAL 90.47 1.223 170.45 1.44 !
4468 E2-9
TAllLE 2 (Sheet 1 of 2)
Contribution to the I.OCA Dose Exposures for a Maximum MSIV Leak Rate of 100 scfh IIntch Nuclear Plant Whole llody Thyroid lleta
Exclusion Area 25.0 300.0 ** '
A) 10 CFR 100 Limit Boundary (2 hr) B) Previous Calculated 1.223 90.47 Doses
- C) Contribution From 0.003 0.047 ,
MSIVs at 100 scfh D) New Calculated Doses 1.226 90.517 Low Population 25.0 300.0 **
A) 10 CFR 100 Limit Zone (30 days)
B) Previous Calculated 1.44 170.45 Doses
- C) Contribution From 0.20 43 14 MSIVs at 100 scfh D) New Calculated Doses 1.64 213.59 Control Room A) GDC-19 5.0 30.0 30/75 (30 days) *** -
B) Previous Calculated 0.066 10.0 1.0 Doses
- C) Contribution From 0.11 5.01 1.54 MSIVs at 100 scfh D) New Calculated Doses 0.176 15.01 2.54 4468 E2-10 l
l
. . .- .. . . _ - - - - _ . .-. -- . . = - --- - - . ._ -. .
l TAlllE 2 (Sheet 2 of 2)
Contribution to the LOCA Dose ICxposures for a Slaximum SISIV 1 cak Rate of 100 sclh l llatch Nuclear Plant Whole Body Thyroid Beta (rem) (rem) (rem)
Technical A) GDC-19 5 30 30/75 Support Center "*
(Filter started at time of B) Previous Calculated 0.0328 2.96 0.996 accident) Doses
- C) Contribution From 0.06 3.01 1.59 MSIVs at 100 scfh D) New Calculated Doses 0.0928 5.97 2.586 Technical A) Previous Calculated 0.0316 21 06* * *
- 0.944 Support Center Doses *
(Filter started 10 min afler B) Contribution From 0.06 3.01 1.59 ,
accident) MSIVs at 100 scfh C) New Calculated Doses 0.0916 24.07 2.534 Technical A) Previous Calculated 0.0286 17.46 0.854 Support Center Doses *
(Filter started 30 min aller B) Contribution From 0.06 3.01 1.59 accident) MSIVs at 100 scfh C) New Calculated Doses 0.08 20.47 2.444
- Recalculated doses from Table 1.
- No limit specified.
- *
- 75 if prior commitment has been made to use protective clothing.
4468 E2-11
Enclosure 3 Edwin I. Hatch Nuclear Plant - Unit 2 i Request to Revise Technical Specifications:
Increase in Allowable MSIV Leakage Rate and Deletion of the MSIV Leakage Control System 10 CFR 50.92 Evaluation t
Proposed Changd This proposed change increases the allowable leak rate specified in Technical Specification !
3.6.1.2.c and the associated Action from 11.5 standard cubic feet per hour (scfh) for any one main steam isolation valve (MSIV) when tested at 28.8 psig to 100 scfh for any one ;
MSIV with a total maximum pathway leakage of 250 scth through all four main steam lines when tested at 28.8 psig. Any MSIV that exceeds 100 scfh will be restored to 11.5 scflt If the total MSIV leakage for all four main steam lines exceeds 250 scfh, the l
. necessary MSIVs will be restored such that the maximum pathway leakage for all four
! main stream lines is no more than 250 scfh.
Basis for Proposed Change 1 Georgia Power Company (GPC) has reviewed the proposed change and determined it does not involve a significant hazards consideration based on the following:
- 1. The change does not involve a significant increase in the probability or consequences ;
of an accident previously evaluated. The proposed amendment does not involve a .
change to structures, components, or systems which would afTect the probability of ,
an accident previously evaluated in the Ilatch Unit 2 Final Safety Analysis Report l (FSAR). It results in acceptable radiological consequences for the design basis loss 1 of coolant accident (LOCA) which was previously evaluated in section 15.1.39 of the FSAR. The MSIV leakage contribution to control room and ofTsite doses are bounded by the LOCA as described in section 15.1.39 of the FSAR. Therefore, the proposed amendment will not significantly increase the consequences of other 1 analyzed accidents.
Plant specific radiological analyses have been performed to assess the efTects of the proposed increase in the allov.able MSIV leak rate in terms of control room, technical l suppon center (TSC), and offsite doses following a postulated design basis LOCA.
4468 E3-1
. . . .- -. ~
,, . s Enclosure 3 ,
10 CFR 50.92 Evaluation f
These analyses utilize the hold-up volumes of the main steam piping and condenser as an alternate method for treating MSIV leakage. The radiological analyses use standard conservative assumptions for the release of source terms consistent with Regulatory Guide 1.3, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Boiling Water Reactors,"
Revision 2, dated April 1974.
The analysis demonstrates dose contributions from the proposed MSIV leakage rate limit of 100 scih per MSIV with a maximum MSIV leakage not to exceed 250 scfh
. through all four main steam lines, results in an acceptable increase to the LOCA doses previously evaluated against the regulatory limits for the offsite, control room, and TSC doses as contained in 10 CFR 100 and 10 CFR 50. Appendix A (General Design Criterion 19). The LOCA doses previously evaluated Oecalculated) are discussed in i section 15.1.39 of the FSAR. The revised LOCA doses are the LOCA doses previously evaluated in the FSAR (recalculated doses to be incorporated into the FSAR) plus the MSIV leakage doses calculated assuming use of the alternate treatment method. Table 2 of Enclosure 2 shows the previously calculated doses and ,
the newly calculated doses.
It is important to note the resulting doses are dominated by the organic iodine fractions which occur because of the conservative source term assumptions used in this analysis. For a total MSIV leakage of 250 scfh through all four main steam lines, more than 80 percent of the offsite, control room, and TSC iodine doses are due to the organic iodine from Regulatory Guide 1.3 source term and organic iodine converted from the elemental iodine deposited in main steam piping systems. If the actual iodine composition from the fuel release (cesium iodine) is used in the calculations essentially all of this organic iodine dose would be eliminated :
The TSC doses due to MSIV leakage are especially conservative. It is not expected i that there will be any radioactive releases to the TSC due to MSIV leakage during the first 30 minutes following a LOCA event since it would take considerable time for the MSIV leakage to travel through the main steam lines and main steam line drain system to the condenser, into the turbine building, and finally to the atmosphere and TSC. Ilowever, it was conservatively assumed the 30-day integrated dose of s-3.01 rem due to a total MSIV leakage of 250 scfh could be received by personnel who start the filter system 30 minutes afler the LOCA occurs and immediately enter ;
the TSC. The dose calculations were made using control room occupancy factors .i specified in SRP 6.4. i 4468 E3-2
l Enclosure 3 10 CFR 50.92 Evaluation l
l
- 2. The proposed change will net create the possibility of a new or different kind of accident from any accident previously analyzed. The BWROG evaluated MSIV leakage performance and concluded MSIV leakage rates up to 200 scfh will not inhibit the capability and isolation performance of the valves to isolate the primary containment. There is no new modification which could impact the MSIV operability. The LOCA has been analyzed using the main steam piping and condenser as a treatment method to process MSIV leakage at the proposed maximum rate of 250 scfh through all four main steam lines. Therefore, the proposed change will not create any new or different kind of accident from any accident previously analyzed m '
the FSAR. ;
- 3. Operation of Plant Ilatch in accordance with the proposed change will not involve a significant reduction in the margin of safety. The allowable leak rate limit specified ;
for the MSIVs is used to quantify a maximum amount of bypass leakage assumed in the LOCA radiological analysis. Results of the analysis are evaluated against the dose requirements contained in 10 CFR 100 for the ofTsite doses and 10 CFR 50, ,
Appendix A (General Design Criterion 19) for the control room and TSC doses.
The margins of safety are not significantly adversely afTected because the dose levels !
remain well below the limits of 10 CFR 100 and General Design Criterion 19.
Therefore, the proposed amendment does not involve a significant reduction in the margin of safety at Plant Hatch.
Proposed Change 2 j This proposed change to delete Technical Specification 3/4.6.1.4 and Bases section 3/4.6.1.4 involves eliminating the MSIV leakage control system (LCS) requirements from l the Technical Specifications. ;
Basis for Proposed Change _2 Georgia Power Company has reviewed the proposed change and determined it does not !
involve a significant hazards consideration based on the following: ;
- 1. The proposed change does not involve a significant increase in the probability or i consequences of an accident previously evaluated. As described in section 6.5 of the l FSAR, the LCS is manually initiated about 20 minutes following a design basis I
l l
4468 E3-3
Enclosure 3 10 CFR 50.92 Evaluation j l
i LOCA. Since the LCS is operated only after an accident has occurred, this proposed amendment has no effect on the probability of an accident. The proposed change results in acceptable radiological consequences of the design basis LOCA previously evaluated in section 15.1.39 of the FSAR. l Plant Hatch has an inherent MSIV leakage treatment capability. GPC proposes to use the main steam line drains and condenser as an alternate to the LCS. GPC will incorporate this alternate method in the Operating Procedures and Emergency Operating Procedures, consistent with the incorporation of the existing leakage ;
control system.
Plant specific radiological analyses have been performed to assess the efTects of MSIV leakage in terms of control room, TSC and offsite doses following a postulated design basis LOCA. These analyses utilize the hold-up volumes of the main steam ,
piping and condenser as an alternate treatment method for the MSIV leakage. The i analysis demonstrates the proposed change results in an acceptable increase in the radiological consequences of a LOCA previously evaluated in the FSAR. Since the ,
MSIV leakage contribution to control room, TSC, and ofTsite doses is bounded by ;
the LOCA as described in section 15.1.39 of the FSAR, the proposed change will not i involve a significant increase in the consequences of an accident previously analyzed. >
- 2. The proposed change does not create the possibility for a new or different kind of accident from any accident previously analyzed. The purpose of the LCS is to reduce j the untreated MSIV leakage when isolation of the primary coolant system and containment are required. Radiological dose contributions due to MSIV leakage are l bounded by a LOCA. The LOCA has been analyzed using the main steam piping and .
condenser as a treatment method to process MSIV leakage at the proposed maximum rate of 100 scfh and 250 scfh total maximum pathway leakage, and determined to be >
within the regulatory requirements. Therefore, the proposed change does not create
. the possibility of a new or difTerent kind of accident.
- 3. The proposed change to delete Technical Specification 3/4.6.1.4 and Bases section 3/4.6.1.4 does not involve a significant reduction in the margin of safety. The ;
intended function of the LCS for treatment of MSIV leakage will be performed by :
using the more efTective alternate path via the main steam drain lines and condenser.
]
This treatment method is effective for treatment of MSIV Icakage over an expanded leakage range. Except for the requirement to assure certain valves are opened to i
l 4468 E3-4 l
l
Enclosure 3 10 CFR 50.92 Evaluation P
establish a proper flow path from the MSIVs to the condenser, the proposed method is passive and does not require any logic controls or interlocks. On the other hand, the LCS consists of complicated logic controls and sensitive equipment which must be maintained at significant cost and radiation exposure. The radiological effects on the margin of safety are discussed above fbr Change 1. The safety significance of the LCS in terms of public risk was addressed in NUREG/CR-4330 which contains the ;
evaluation for climinating the LCS and disabling the systems currently installed at BWRs. The conclusion was that the increased public risk is less than 1 percent, ,
Therefore, the proposed change does not involve a significant reduction in the margin of safety at Plant Hatch.
I l
l r
l 4468 E3-5
-- - . . . -. - -- __