ML20153E551

Responds to NRC 871022 Request for Addl Info Re Proposed Amend 111,Rev 1 to License DPR-54 Concerning Tech Spec Table 3.6-1,adding 16 Valves & Increasing Max Closure Time of Valves in Direct Flow Path from Containment to Environ

ML20153E551
Person / Time
Site:	Rancho Seco
Issue date:	08/31/1988
From:	Keuter D SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20153E555	List: ML20153E551 ML20153E648
References
NUDOCS 8809060348
Download: ML20153E551 (10)

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$SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT C 62013 Street. P.o. Box 15830, Sacramento CA 95852 1830 (916) 452 3211 AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA AGM/NPP 88-457 Directcr of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission

1 ail Station PI-137 Washington, DC 20555 Docket No. 50-312 Rancho Seco Nuclear Generating Station License No. DPR-54 PROPOSED AMENDHENT NO. 111, REV. 1, SUPPLEMENT 1

Reference:

NRC Request for Additional Information (G. Kalman to G. C. Andognini, dated October 22, 1987)

Dear Sir:

The referenced letter presented seven items for which additional information was requested concerning Proposed Amendment No. 111, Rev. 1. This Proposed Amendment added 16 valves to Table 3.6-1 of the Rancho Seco Technical Specifications, and increased to 25 seconds the maximum closure time of valves in a direct flow path from the containment atmosphere to the environment.

Attachment I is the District's response to the NRC request for additional information. Attachment II consists of three revised pages of the Technical Specifications. Page 3-40b corrects the margin change marks which erroneously designated the changes for Proposed Amendment No. 161 rather than for Proposed Amendment No. 111. The changes to Table 3.6-1, and the reasons for the changes, are as follows:

1. The four Reactor Building purge valves have been reinserted. These valves may be open during refueling, but automatically close on a high radiation signal such as could occur in a fuel handling accident. The maximum closure time of these four valves has been increased to 25 seconds. This increase is based on the analysis discussed in Item 6 of Attactnent I, and is consistent with the N maximum closure time of the other direct flow path containment g isolation valves.

hC0 2. The four hydrogen recombiner isolation valves have been deleted.

cdo These valves (HV-53620, 21, 22, and 23) were installed in accordance co w with the District's earlier plan that hydrogen recombiners external vu to the containment building would be used at Rancho Seco for compli-

$8 ance with 10 CFR 50.44(b)(3). The District subsequently elected to

$< install internal hydrogen recombiners, for which purge valves are S unnecessary. Amendment No. 95, approved February 12, 1988, added O$ these internal hydrogen recombiners to the Technical Specifications.

(D A A Presently, the four isolation valves for external hydrogen recombiners are locked closed with the external valves blind-flanged, and the f power removed from the motor operators of these hur valves, p 08l RANCHO SECO NUCLEAR GENERATING STATloN D 1444o Ten Cities Road, Herald, CA 95638 9799; (209) 333 2935 l g

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. 7 AGH/NPP 88-457 Director of NRR 3. The maximum closure time of the reactor coolant pump seal return (SFV-24004) has been decreased from 71 seconds to <60 seconds. This reduction in closure time is in accord with B&H recommendations, and complies with Standard Review Plan 6.2.4 which states: "In general, valve closure times should be less than one minute."

The Safety Analysis and No Significant Hazards Consideration remain unchanged from those previously submitted with Proposed Amendment No. 111, Rev. 1.

Pursuant to 10 CFR 59.91(b)(1), the Radiological Health Branch of the  ;

California State Department of Health Services has been informed of this Proposed Amendment by mailed copy of this submittal.

The District has determined that no license fees are required with this revision since a check for $150.00 was sent to the Commission with the initial submittal of Proposed Amendment No. 111.

Members of your staff with questions requiring additional information or clarification may contact Mr. Robert Roehler at (209) 333-2935, extension 4918.

State of California SS County of Sacramento Hilliam E. Kemper, being first duly sworn, deposes and says: that he is Manager, Operations of Sacramento Municipal Utility District (SHUD), the licensee herein; that he has been authorized per Interoffice Memorandum AGM/NPP 88-472 to sign for the AGH, Nuclear Power Production; that he has executed the foregoing document; that the statements made in this document are true and correct to the best of his knowledge, information, and belief, and that he is authorized to execute this document on behalf of said licensee.

- C p M n R. Keuter/

Assistant General Manager, Nuclear Power Production Subscribed and affirmed to before me on this J f day of[d N Ei ,1988.

... omci AL SAL"W Esther Hughes aG;) i

' ESTHER H. HUGHES h Notary Public

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. l AGM/NPP 88-457 l Director of NRR Attachments cc w/atch:

A. D'Angelo, NRC, Rancho Seco i J. B. Martin, NRC, Region V (2)

MIPC (2) i

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INPO J. S. McGurk, State of California i f

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ATTACHMENT I RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION

7 ATTACHMENT I District Response to Request for Additional Information The District's response to the Request for Additional Information (RAI) received from the NRC on October 22, 1987 is presented below. For review convenience, each RAI is presented first, followed by the District's response.

References:

1. SHUD transmittal to NRC of Proposed Amendment No. 111, Rev. O, dated February 2, 1986

2. SMUD transmittal to NRC of Proposed Amendment No.111, Rev.1, dated May 14, 1987

1. In Ref. 1 SMUD proposed to: (1) delete 4 reactor building purge valves from TS Table 3.6-1, p. 3-4G; (2) add 16 additional containment isolation valves to 1S Table 3.61; and (3) increase the maximum closure time listed in Table 3.6-1 for 34 of the containment isolation valves to 25 seconds.

Ref. 1 (Attachment II, p. 5) stated that: "The increase in closure time to 25 seconds of selected isolation valves has, by analysis, been determined to result in offsite doses which remain within 10 CFR 100.11 limits." Identify the "analysis" referred to in the preceding sentence, and submit a copy if SMUD h?.s not already submitted it.

Runoni.e:

The analysis referred to on Page 5 of Att&chment II to Reference 1 is the summation of the values presented in SMUD calculation Z-ZZZ-M1304, Revision 1 transmitted in Reference 2, Attachment IV and in USAR Table 14.3-3.

2. P ihough the Updated Safety Analysis Report (USAR, Ch. 14) for Rancho Seco lists the radiological consequences for many accidents, it appears that SMUD (Ref. 2. Attachment IV) has considered the radiological impacts from only the loss-of-coolant-accident (LOCA) during containment purge. For each of the accidents listed in the FSAR, state whether the radiological sequences of that accident will be increased by the proposed TS changes.

For those accidents whose radiological consequences will not be increased by the proposed TS changes, briefly state the basis for SMUD's position

, that the consequences will not be increased. For those accidents whose radiological consequences will be increased, provide a revised accident

. analysis. The revised accident analyses should identify the impact that I each of the changes in Table 3.6-1 will have on the estimated doses. If I it is SMUD's position that the addition or deletion of a particular valve, or an increr.se In the closure time, clearly has a negligible impact and it is not necessary to reestimate the doses for that particular valve, then briefly state the basis for SMUD's position that the impact is negligible.

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7 Rupnie:

USAR Chapter 14 lists the radiological cor. sequences for a LOCA as well as other accidents. Those accidents other than the LOCA are:

a. Loss of All A-C Power

b. Steam Line Failure

c. Steam Generator Tube Fa'. lure

d. Fuel Handling Accident

e. Letdown Line Rupture

f. Control Rod Ejection Accident The impact of the proposed Technical Specification changes on each of these events follows:

Loss of All A-C Power -- The source of the radiological release in this event is the safety valve. These valves are located outside the Reactor Building; therefore, the proposed change has no impact.

Steam Line Failure -- Two steam line failures are presented in the USAR.

For the failure located outside containment, the proposed change has no impact on the radiological consequences. For the failure located inside containment, the USAR states that the loses resulting from a steam line failure are negligible. The source of the radiological release is the 0.1% vol. per day leakage from the Reactor Building for the duration of the event. Based on the relatively low radiological source term for the secondary side coupled with the rapid isolation of the Reactor Building (less than 1 minute), anu the acceptability per Regulatory Guide 1.4 of lowering the leakage rate to 0.05% vol. per day after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, adequate conservatism exists in the current analysis to bound the effect of the proposed change.

Steam Generator Tube Failure -- As in the Loss of All A-C Power event discussed above, the source of the radiological release for this event is the safety valve. Since these valves are located outside the Reactor Building, the proposed change has no impact for this event.

Fuel Handling Accident -- The radiological consequences for this event are presented in response to RAI 6.

Letdown Line Rupture -- Similar to the Loss of All A-C Power event discussed above, the source of the radiological release (a oroken letdown line in Auxiliary Building) is located outside the Reactor Building; therefore, the proposed change has no impact on this event.

Control Rod Ejection Accident -- As stated in USAR paragraph R t.2.4.6, the maximum diameter hole size resulting from a rod ejection is approximately 2.76 inches. This is smaller than the smallest rupture size evaluated in the LOCA analysis. This results in a lower rate of energy input into the Reactor Building and subsequently a lower Reactor Building pressure. The analysis presented in USAR paragraph 14.3.7 assumes a 0.1%

vol. per day leakage for the duration of the accident. This leakage value  !

is consistent with large break LOCA pressures and is therefore highly conservative, fince the Reactor Building pressures are low and the establishment or containment intagrity occurs within a very short time (less than 1 minute), the existing analysis provides sufficient conservatism.

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' 3. R'ef. 1 (Attach ent II, p. 6) states that: "Calculational results of the doses to the Low Population Zone (LPZ) and Exclusion Area Boundary (EAB) due to the MHA are provided in Enclosure I." Although we have infor-ally.

obtained a copy of page 4 of Enclosure I, we have not received the entire Enclosure I. He also note that the estimated doses in Enclosure I (p. 4) are different than the values in Ref. 2, Attachment IV, p. 4. Provide a copy of Enclosure I, and clarify the meaning of these values.

Response

Calculation Z-ZZZ-M1304, Revision 1 determines the short term dose due to the release of containment atmosphere during the time required to close the containment isolation valves. In contrast, USAR Table 14.3-3 presents the doses from the HHA source term releases due to containment leakage from time O to 30 days. The results of this "analysis" are presented as Enclosure I "Maximum Integrated Doses in Rems due to Environmental Releases During an MHA" to this attachment.

4. The Branch Technical Position CSB 6-4 (Ref. 3, p. 6.2.4-15, Section B.1.f) states that: "Purge system isolation valve closure times, including instrumentation delays, should not exceed five seconds, to facilitate compliance with 10 CFR 100 regarding radiological consequences." Although SMUD states that the proposed changes will not have any adverse impacts on plant safety (e.g., see Ref. 2, Attachment I, p. 2), presumably not increasing the valve closure times will not have any adverse impacts on plant safety. State why the containment isolation va:.ves at Rancho Seco should be allowed to have closure timer, much greater (i.e., up to 25 seconds) than is provided for in the Branch Technical Position. Provide a justification for increasing the maximum closure time from the present value in the TS to the proposed value for each valve or group of valves whose maximum closure time exceeds or will exceed 5 seconds.

Eftsponie:

Th: District executed a design change to implement a Reactor Building pressure equalization system (mini-purge). Although the original design did not strictly adhere to the guidance provided in BTP CSB 6-4, an analysis has been performed that supports the design criteria by meeting the requirements of 10 CFR 100.11.

The current Technical Specification calls for closure times of 9 to 15 seconds for the equalization valves. (It should be noted that the current Technical Specification does not adhere to the BTP CSB 6-4 guidance.)

Surveillance testing of these valves indicated that the closure times required by the Technical Specification were marginally met. To allow the installed equalization valves to consistently pass the su% eillance testing and to minimize wear on the valve and its operator, an increase in .

the maximum allowable closure time is proposed by this Technical l Specification amendment. As discussed above, an analysis has been  !

performed to evaluate the effects of this change. The results justify the 1 radiological acceptability of the requested amendment. I i

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7 For valves other than sini-purge valves, Standard Review Plan (SRP) 6.2.4 provides a closure time acceptance criteria of less than 1 minute, regardless of valve size. Calculation Z-ZZZ-M1304, Revision 1, evaluated the doses due to leakage during the establishment of containment integrity.

This analysis includes both the equalization (mini-purge) valves as well as other pertinent containment isolation valves. The results of this analysis demonstrate compliance with the SRP by meeting the requirements of 10 CFR 100.

5. SMUD states (see Ref.1, p.1) that their position is that "the modeling of the release of radioactivity to the containment environment need not be instantaneous, as would be assumed with the use of TID source terms during the first few minutes of a LOCA." SHUD referenced Section B.5.a of the Branch Technical Position (Ref. 3) as supporting their position. Howevv.

as noted above, some of the proposed valve closure times exceeded the guidance contained in Section B.1.f. and thus the use of a TID source term is appropriate.

Responia:

The District restates its position that modeling of the release of radioactivity 'o the containment environment need not be instantaneous.

The District considers that Section B.5.a of BTP 6-4 supports this position. He have reviewed Section B.1.f of BTP 6-4 and agree that while i the valve closure time is a sound recommendation, it in no way implies that valve closure times in excess of this recommendation require TID source term modeling. As stated in the bases to Technical Specification 3.6, "When containment integrity is established, the limits of 10 CFR 100 will not be av aa4d :hould the maximum hypothetical accident occur," and that "Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for LOCA."

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6. In a telephone conversation on September 1,1987, the staff indicated to the licensee that the deletion of the 4 purge valves from TS Table 3.6-1 was not acceptable, and the licensee agreed to put these valves back in the table but with a longer closure time. Provide justifications for the longer closure time for the 4 purge valves.

Respante:

An analysis was performed to determine the maximum acceptable valve closure time for the containment purge valves. Technical Specification 3.6 limits the opening of the containment purge valves to cold shutdown or refueling.

Given this limited operation, a fuel handling accident was determined to be the design basis accident for the containment purge valve isolation times. Regulatory Guide 1.25 and SRP 15.7.4 assumptions were used in the analysis. SRP 15.7.4 also provided the acceptance criteria of 75 Rem thyroid dose and 6 Rem whole body dose.

The analysis described above was perforEed assuming no isolation of the containment purge valves. The results of the analysis were as follows:

Dose in Rems EAB LPZ Thyroid dose 7.19 2.54E-1 Whole body dose 4.77E-1 1.68E-2 Beta skin dose 7.70E-1 2.71E-2 These results are well within the acceptance criteria of SRP 15.7.4 as well as 10 CFR 100.11. SRP 6.2.4 guidelines recommend that containment isolation valves have a closure time less than 1 minute. Based on the above analysis, any containment purge valve isolation time of less than 1 minute is acceptable.

7. The addition of 16 containment isolation valves to TS Table 3.6-1 is acceptable provided that all of the valves satisfy the applicable regulatory requirements for containment isolation valves. In Reference 1 (Attachment II, p. 4 and 5), the licensee indicated that the hydrogen monitor isolation valves, hydrogen recombiner isolation valves and reactor building hydrogen purge valves (HV-53617, HV-53618' met the requirements of Standard Review Plan 6.2.4.11.6.r, and GDC 66 of 10 CFR 50, Appendix A.

The staff found that the GDC 66 was the wrong reference criterion. It is not clear whether the reactor building hydrogen purge valves (SFV-53615 and SFV-53616) meet the applicable requirements. The licensee indicated in a telephone conversation on September I, 1987 that GDC 66 was a typographical error, and the two valves (SFV-53615 and SFV-53616) met the specified requirements of GOC 56. Further, the licensee committed to clarify this in a revised submittal. Provide the written clarification.

Response

Attachment II, Pages 3 and 4 to the District's submittal of Technical Specification Amendment No. Ill, Revision 1, discuss the "Effects on Safety functions" of the Reactor Building Hydrogen Purge containment isolation valves. It was stated that these valves and the associated containment isolation systems meet General Design Criteria 66 of 10 CFR 50, Anpendix A. The statement should have indicated that GOC 56 was the applicable design criteria.

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. 1 ENCLOSURE I Maximum Integrated Doses in Rems f due to Environmental Releases l During an MHA Integrated 2-Hour Exposure at 0.4 Miles (EAB)

Maximum Hvoothetical Accident Thyroid Whole Body Containment Leakage (0-30 Day) 137 3.60 Leakage during establishment of containment integrity 22.3 .01 Total 169.3 3.61 10 CFR 100 Guidelines 300.0 25.0 Integrated 30 Day Exposure at 0.4 Miles (LPZ)

, Faximum Hvoothetical Accident Thyroid Whole Body Containment Leakage (0-30 Day) 8.40 0.299 Leakage during establishment of containment integrity 1.14 neglig Total 9.54 0.299 1C CFR 100 Guidelines 300.0 25.0

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