ML20078K468

Safety Evaluation Supporting Amend 203 to License DPR-49

ML20078K468
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Site:	Duane Arnold
Issue date:	11/17/1994
From:	Office of Nuclear Reactor Regulation
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y-t UNITED STATES 3

E NUCLEAR REGULATORY COMMISSION f

WASHINGTON D.C. 206M 0001 3

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0. 203 TO FACILITY OPERATING LICENSE NO. DPR-49 IES UTILITIES INC.

CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DUANE ARN0LD ENERGY CENTER DOCKET NO. 50-331

1.0 INTRODUCTION

In 1991, IES Utilities Inc. (the licensee), formerly known as Iowa ElectHc Light and Power Company, performed its own independent review of the Du ne Arnold Energy Center (DAEC) Technical Specifications (TS) as part of a self-initiated TS improvement program. A portion of the program included comparison of the Cuane Arnold TS with TS from similar plants, the General Electric Standard TS (NUREG-1202, July 1986), and the draft improved Standard l

TS (NUREG-1433).

Based on that comparison, the licensee, by letter dated December 31, 1992, proposed changes to TS Section 3.6, " Primary System Boundary." Subsequent to that submittal, the licensee discovered some erroneous references and inconsistencies in their proposed changes.

By letter dated June 4,1993, the licensee corrected the errors and inconsistencies in the original submittal.

The June 4, 1993, submittal, superseded the original submittal and this evaluation is based solely on the June 1993 submittal and TS Bases changes in the May 6, 1994, submittal.

The submittal dated May 6, 1994, repeated a TS change (deletion of the inservice inspection interval start date) which was in the June 4, 1993, application, and provided 3/4.6 Bases changes.

The request for amendment proposes several changes to the DAEC Technical Specification (TS) Section 1.0, " DEFINITIONS," and a number of changes to Limiting Conditions of Operation (LCO) and Surveillance' Requirements (SR) in TS Section 3/4.6, " PRIMARY SYSTEM B0UNDARY." These changes include appropriate revisions of the corresponding TS Bases Sections.

The TS 3/4.6 LC0 and SR Subsections that are affected are listed below:

TS 3/4.6.A. - Thermal and Pressurization Limitations TS 3/4.6.B. - Coolant Chemistry TS 3/4.6.C. - Coolant Leakage TS 3/4.6.D. - Safety and Relief Valves TS 3/4.6.E. - Jet Pumps TS 3/4.6.F. - Jet Pump Flow Mismatch TS 3/4,6.G. - Structural Integrity TS 3/4.6.H. - Shock Suppressors Y Nbbo333 i

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I The proposed changes to Section 3/4.6 are intended to clarify existing Limiting Conditions for Operation (LCO), Surveillance Requirements (SRs), add specific shutdown requirements and provide consistency with the rest of the plant TS and Standard TS. The proposed changes are plant specific in nature in that they clarify the existing TS and provide consistency in the overall Duane Arnold TS.

This amendment request also involves a substantial number of simple editorial changes to the TS 3/4.6.

These changes include renumbering, capitalization of defined terms and replacing words to be consistent with the Standard TS. The existing wording is sometimes vague and misleading and could lead to misinterp atation.

Index page v also corrected the page location of Figure i

1.0-1.

The ;;taff, therefore, concludes that these editorial changes are acceptable th:oughout TS Section 3/4.6.C, as they are more specific and l

clearer than the existing TS and they would not affect the operation of the DAEC facility or adversely impact safety. These editorial changes will not be listed in this SER since they do not change the operation of the plant.

l 2.0 EVALUATION I

2.1 TS Section 1.0 - DEFINITIONS 3

The DAEC Amendment request submittal proposes to add definitions for IDENTIFIED LEAKAGE, TOTAL LEAKAGE, UNIDENTIFIED LEAKAGE, AND DOSE EQUIVALENT 4

1 1-131 to Section 1.0 of the DAEC TS, as proposed below:

3 Definition 41, IDENTIFIED LEAKAGE

" IDENTIFIED LEAKAGE shall be:

a.

Leakage into collection system, such as pump seal or valve i

packing leaks, that is captured and conducted to a sump or collecting tank, or b.

Leakage into the containment atmosphere from sources that are both specifically located and known not to interfere l

with the operation of leakage detection systems."

Definition 42. TOTAL LEAKAGE j

" TOTAL LEAKAGE is the sum of IDENTIFIED LEAKAGE and UNIDENTIFIED LEAKAGE "

Definition 43, UNIDENTIFIED LEAKAGE

" UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE."

Definition 44, DOSE EQUIVALENT I-131

" DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram (ml), which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, I-133, 1-134, and I-135 actually present.

The thyroid dose factors used for this calculation shall be those listed in Tab!e III

,.-.,...~

. of TID-14844, " Calculation of Distance Factors for Power and Test i

Reactor Sites."

The licensee proposes to add Definition 44, DOSE EQUlVAL ENT I-131, to provide clarity and consistency with LC0 3.6.8., " Coolant Chemistry," and the corresponding Standard TS definition.

This definition is acceptable to the staff.

The licensee proposes to add Definition 41, IDENTIFIED LEAKAGE, Definition 42, TOTAL LEAKAGE and Definition 43, UNIDENTIFIED LEAKAGE, to provide clarity and consistency with LC0 3.6.C., " Coolant Leakage," and the corresponding Standard TS definition.

Proposed Definitions 41, 42, and 43 are appropriate for LCO 3.6.C. as currently licensed, and for the proposed changes to LCO 3.6.C.

The staff, therefore, finds the addition of definitions 41 through 44 acceptable.

2.2 TS 3/4.G.A. - Thermal and Pressurization Limitations The licensee has proposed to revise TS Section 3.6.A.2 to delete from the LC0 a commitment from the licensee to update TS Figure 3.6-1, Pressure versus Minimum Temperature Valid to Sixteen Full Power Years, per Appendix G of 10 CFR Part 50, prior to the expiration of 16 full power years, and to relocate the commitment to the Bases Section of the TS. This Figure gives the Pressure

- Temperature Limit Curves for the DAEC reactor vessel.

The staff approved figure 3.6-1 on August 12, 1991.

This commitment does not affect any surveillance requirements on the LCO, or provide any additional information to assist in the operation of the plant or in mitigating any accidents which could potentially occur.

The staff has determined that this change is acceptable.

The licensee has proposed to revise TS Section 3.6.A.3 to identify the head flange and the shell adjacent to the head flange as the locations where temperature readings are to be taken before the licensee may place the reactor vessel head bolting studs in tension.

These locations are acceptable to the staff and the change is acceptable.

The licensee has proposed to add Action Statement, TS Section 3.6.A.4, which requires the licensee to take action if any of the pressure - temperature (P/T) limits required by the TS Section 3.6.A.l.,

2., or 3. are exceeded:

1.

Average heatup/cooldown rate not to exceed 100 *F/hr 2.

Reactor head required to be vented and no power operation allowed unless the reactor vessel temperature is greater or equal to Curve C in Figure 3.6-1 3.

Reactor Vessel Studs shall not be placed in tension unless the temperature of the head flange and shell adjacent to the head flange are greater than 74 'F.

. Should any of these limits be exceeded, the licensee is required to restore the temperature or pressure to the acceptable level within 30 minutes of discovery, and to perform an evaluation of the out-of-limit condition on the structural integrity of the Reactor Coolant System (RCS), in order to determine whether the RCS remains acceptable for continued operation.

If these conditions cannot be met, the licensee will be required to be in Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Cold Shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The staff has reviewed the licensee's proposed Action Statemcat and has determined that it is more stringent than the comparable Action Statement in the Standard Technical Specifications, Therefore, this Accion Statement is acceptable to the staff.

The licensee has proposed to apply the RUN, STARTUP, HOT SHUTDOWN, COLD SHUT 00WN, and REFUELING Modes of operation to TS Section 3.6.A.5, in regard to operation of the recirculation pcmp.

These Modes of operation are consistent with the OPERATIONAL MODES in the Definitions Section of the TS and are acceptable to the staff.

The licensee has proposed to revise TS SR 4.6.A.2 to require that removal of vessel integrity specimens be done in accordance with 10 CFR Part 50 Appendix H.

The SR also requires that the results of the specimen fluence determinations will be used to update TS Figure 3.6-1.

This SR is in agreement with the requirements of Generic Letter 91-01, " Removal of the Schedule for the Withdrawal of Reactor Vessel Material Specimens from Technical Specifications." This revision to TS SR 4.6.A.2, therefore, is acceptable to the staff.

2.3 TS 3/4,6.6. - Reactor Coolant Chemistrv The licensee proposed to revise the TS to clarify LC0 Section 3.6.B.1 by revising the present LC0 into three different sections.

Section 3.6.B.I.a will be revised to state:

. "With the reactor critical, the specific activity of the primary coolant shall be less than or equal to 1.2 Ci/ml DOSE EQUIVALENT I-131."

Section 3.6.B.l.b will be revised to state:

"When in the RUN, STARTUP, or HOT SHUTDOWN MODE, the specific activity of the primary coolant can be greater than 1.2 yCi/mi DOSE EQUIVALENT I-131 for a maximum of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided that the DOSE EQUIVALENT I-131 activity does not exceed 12.0 Ci/ml during this time.

The reactor shall not be operated more than 5 percent of its yearly power operation under this exception for equilibrium activity limit."

Section 3.6.B.I.c will be revised to state:

. "If the specific activity of the primary coolant is greater than 12.0 Ci/ml DOSE EQUIVALENT I-131, the reactor shall be shutdown and the Main Steam Line Isolation Valves shall be closed immediately."

f The footnote, "That concentration of I-131 which alone would produce the same thyroid dose as the quantity and isotopic mixture actually present." is no longer needed and will be deleted.

The surveillance requirements (SRs) for Section 4.6.8 will be revised by placing the information in proposed tables or by formatting in accordance with the guidance provided by the standard TS.

SR 4.6.B.1.a will be revised to state:

"The specific activity of the reactor coolant shall be demonstrated to be within limits by performance of the sampling and analysis program of Table 4.6.B.1-1."

This SR will be revised and incorporated into proposed Table 4.6.8.1-1.

Placing this information into a tabular format presents it in a clear and concise manner.

The sample frequency has changed from 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

This proposed change is consistent with the guidance in the Standard TS.

SR 4.6.B.I.b, SR 4.6.B.I.c and SR 4.6.B.l.d will be deleted and incorporated into the proposed Table 4.6.B.1-1.

Placing this information into a tabular format presents it in a clear and concise manner.

This proposed change is consistent with the guidance in the Standard TS.

SR 4.6.8.1.e will be deleted and the sample requirements incorporated into the proposed Table 4.6.8.1-1.

Placing this information into a tabular format presents it in a clear and concise manner.

In addition, the requirement pertaining to the sampling during off hours is being deleted.

SR 4.6.B.I.f will be deleted and the sampling requirements revised and incorporated into the Table 4.6.B.1-1.

The requirements in this Table are consistent with the guidance provided in the Standard TS.

SR 4.6.8.1 9 will be deleted and the surveillance requirements and frequencies for sampling the Reactor Coolant System included in the proposed Table 4.6.8.1-1.

SR 4.6.B.I.h will be revised to SR 4.6.B.I.b and state:

"Whenever the DOSE EQUIVALENT I-131 exceeds 0.6 pCi/ml, notify the USNRC as specified by 6,ll.l.h."

SR 4.6.B.I.e, referenced in the old SR 4.6.B.l.h will no longer exist; it will be replaced by informatien in Table 4.6.B.1-1.

The words "(50%

of the equilibrium value)" will also be deleted.

Again, the Table provides adequate guidance as to the requirements.

The unit of measurement "gm" will be replaced with "ml."

DAEC chemistry uses milliliters as a unit of measure.

This change in units will not change the intent or any limits in the existing TS.

The word "by" replaces the word "in," as an editorial change.

The_ existing TSs were written when DAEC used gross iodine radiochemistry and sodium iodide detectors.

Currently, DAEC is using the Dose Equivalent Methodology and no longer uses sodium iodide detectors.

The Dose Equivalent Methodology provides a more quantitative and accurate analysis through the use of the isotopic analysis.

The current Technical Specification calls for an isotopic analysis, as well as a gross measurement on each sample.

The LCOs

. and SRs will be revised by placing the information in proposed Tables or by formatting in accordance with the guidance provided by the Standard TS.

These revisions improve clarity, are consistent with current industry practices, and provide additional guidance not specifically stated in the existing DAEC Technical Specification.

The proposed revision will also enhance the DAEC primary coolant chemistry surveillance program.

The staff finds that the licensee's proposed Technical Specifications change, incorporating an improved isotopic analysis for dose equivalent Iodine-131 in the primary coolant system required by Technical Specification 3/4.6.B.1, is more conservative than the current Technical Specifications and is consistent with the Standard Technical Specifications for General Electric Boiling Water Reactors, NUREG-0123, Revision 3, and therefore acceptable.

The proposed changes move the existing LC0 limits (LC0 3.6.B.2.a.,

b., and d.,

and LC0 3.6.B.3.a.) for conductivity, chloride content, and pH, applicable during operation, shutdown and refueling conditions. to TS Table 3.6.8.2-1.

The limits proposed in Table 3.6.B.2-1 are equivalent or more conservative than the limits found in the existing LCOs 3.6.8.2.a.,

b., and d.,

and are acceptable to the staff. Under conditions when the unit is in the RUN Mode, the licensee is required to be in STARTUP Mode within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, should any of the applicable limits in Table 3.6.8.2-1 be exceeded for mnre than 72 continuous hours or more than 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> / year cumulatively.

These changes are acceptable to the staff.

The proposed change also incorporates the shutdown requirements, during RUN or STARTUP MODE conditions, when chemistry or activity limits or surveillances are determined to be exceeded.

The existing requirements on shutdown simply require the licensee to perform an orderly shutdown of the unit, without specifying the time in which shutdown is to be achieved.

The proposed new shutdown action statements would require the licensee to be in HOT SHUTOOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after determining that a chemistry or activity limit or surveillance was exceeded, and be in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if the adverse situation was not corrected.

The new shutdown requirements would be followed under the following conditions:

1.

In the RUN Mode, if the conductivity, as measured at 25 *C, exceeds 10.0 pmhos/cm, or if the chloride content exceeds 500 ppb (i.e., maximum limits which require immediate shutdown if they are exceeded during RUN Mode conditions).

2.

In STARTUP or HOT STANDBY Modes, if the applicable limits in Table 3.6.B.2-1 are exceeded.

The new chemistry shutdown requirements are more stringent than the comparable Action Statement in the NUREG-1202 (H0PE CREEK STANDARD TECHNICAL SPECIFICATIONS), and are therefore acceptable to the staff.

The proposed change also requires a water sample to be taken and reactor coolant chloride and pH analyses performed every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, whenever the conductivity exceeds the limits in Table 3.6.B.2-1.

These changes are more stringent than previous requirements and are acceptable to the staff.

1

. The proposed change revises the requirements for continuous conductivity monitoring found in existing LC0 3.6.8.3.b, and SRs 4.6.B.3.a. & 4.6.8.3.b.,

and has moved them into LC0 3.6.B.2.a.3 and SRs 4.6.B.2.e. and 4.6.B.2.f.

The revisions make the following changes to the continuous conductivity requirements:

1.

Reference to the locations of the continuous conductivity monitors would be moved to the TS Bases Section.

2.

Channel checks of the continuous conductivity would be required once every 7 days.

3.

In-line conductivity flow cells would be required to be installed and conductivity measurements taken when all three continuous conductivity monitors are determined to be inoperable, as opposed to when just one continuous conductivity monitor was inoperable.

The staff informed the licensee, as conveyed during the staff's conference call to IES on March 9, 1994, that the change in the req"irements for installation of in-line flow cells appears to be in a non-conservative direction. The licensee forwarded information on March 14, 1994, to address the staff's question in regard to this matter.

The licensee's information indicates that the actions required by the Plant Chemistry Procedure 2.13, Surveillance Test Procedure (STP) 468003, and Chemistry Form 103 will be sufficient to account for the change in the requirement.

The licensee will still be required by TS to maintain continuous conductivity monitoring, or else by the TS and the licensee's Chemistry Program to take appropriate corrective measures and actions when such monitoring is unavailable.

In addition, moving the reference of the continuous conductivity monitor locations to the Bases Section does not affect operation or safety of the facility. Therefore, the staff concludes that the change in the requirement will not result in a significant reduction in the margin of safety for the plant, significantly increase the probability of an accident or malfunction of equipment important to safety, or create a new or different kind of accident from any accident previously evaluated.

This change is, therefore, acceptable to the staff.

2.4 TS 3/4.6.C. - Coolant Leakaae The proposed change revises the applicability portion of LCO 3.6.C.l.,

" Coolant Leakage," from "Any time irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212 'F," to "When in RUN, STARTUP or H0T SHUIDOWN MODE." These changes are consistent with the MODES of OPERATION in the Definitions Section of the TS and are acceptable to the staff.

The existing TS LC0 3.6.C.2 (renumbered 3.6.C.3) provides neither a reference nor specific requirements that define sump system operability.

The proposed TS 3.6.C.3 references the applicable section of the TS Table 3.2-E which i

defines sump system operability.

This proposed change also adds to the clarity of the TS and is, therefore, acceptable.

. The licensee is currently required by existing LCO 3.6.C.3 to commence an orderly shutdown of the reactor to cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of determining that any of the leakage limits have been exceeded, in this amendment request, the licensee's June 4,1993, application proposed revising the Action Statement as follows:

1.

"With the conditions in Specifications 3.6.C.1.a, b, or c not met, reduce the leakage rate to within the limits within 4 h w rs, or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />."

The licensee proposes making the change to the Action Statement on the basis that the change would provide clarity and consistency with the Standard TS.

One of the original, proposed changes to the Action Statement was the addition of a 4-hour period which would allow the licensee to attempt bringing the leakage within proper limits, prior to commencing a shutdown of the reactor.

Members of the staff informed the licensee during a conference call on February 7,1994, that the additional 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> represented a reduction in the requirements.

The staff based its assessment of the additional 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> on the following points:

1.

The majority of licensees are prohibited by their TS to operate any unit with an identified Reactor Coolant Pressure Boundary (RCPB) leak.

The 0 gpm RCPB leakage limit is consistent with the corresponding Specifications in the Hope Creek Standard TS (NUREG-1202) and the BWR-5 Standard TS (NUREG-0123).

2.

By default, RCPB leakage at DAEC falls under the scope of TS Definition 43, " Unidentified Leakage." Therefore, under the current licensing basis, the licensee is allowed to operate the unit with a RCPB coolant loss of up to 5 gpm, and still not be required to shut down the reactor.

This is less stringent than the industry norm. Although 5 gpm is very small in comparison to the total core inventory, a leak of this sort would still place the reactor in an extremely slow loss-of-coolant transient.

3.

The adiitional 4-hour grace period to bring the leakage in line is a reduction, since in the case of identified RCPB leakagt, the licensee is already allowed by its licensing basis to operate at power even with a 5 gpm RCPB leak in effect.

Furthermore, the additional 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not consistent with the Action Statement in NUREG-1202, which states that, "With any Pressure Boendary leakage, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTOOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />."

The licensee informed the staff, during the conversation of February 7,1994, I

that it would withdraw the request for the additional 4-hour grace period.

l The licensee confirmed this by sending a revised LC0 3.6.C.2 to the staff on l

9-a February 14, 1994.

LCO 3.6.C.2. will now read, "With the conditions in Specifications 3.6.C.I.a, b, or c above not met, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />." This revision to proposed LC0 3.6.C.2., as written in the previous sentence, is consistent with the corresponding Hope Creek Standard TS Action Statement and is acceptable to the staff.

TS LCOs 3.6.C.4 and 3.6.C.5 will be added to identify specific actions to take in the event that the sump system is inoperable or if the sump system and the air sampling system are both inoperable.

In the event that the sump system is inoperable, the air sampling system is verified operable and the sump system must be restored to an operable status within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With both the sump system and air sampling system inoperable, the licensee has 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore one system to an operable status, before initiating a shutdown.

Proposed TS LC0 3.6.C.4 is basically a rewrite of the existing LCO 3.6.C.3 with the required actions to be taken more specifically described for clarification and to be consistent with other TS LCO requirements.

The proposed TS LC0 3.6.C.5 will be added to cover a condition not previously addressed in the TS, i.e., the inoperability of the air sampling system, when it is required to be operable.

The existing TS cannot readily be interpreted because they are rather vague and do not adequately address actions to be taken under certain circumstances.

Therefore, the staff concludes that the proposed changes are necessary for clarity and are also consistent with leakage detection system LCOs in the Standard TS and in the TS of other BWRs and are, therefore, acceptable.

2.5 TS 3/4.6,0 - Safety and Relief Valves (SRVsl The existing TS LCO 3.6,0.1 will be revised by adding a note to state that SRVs which perform an Automatic Depressurization System (ADS) function must also satisfy OPERABILITY requirements as specified in TS 3.5.F.

Existing TS LC0 3.6.D.2.a and b. will be revised to clarify the LC0 requirements in the event that the safety function of relief valve (s) becc inoperable. Also, existing LC0 TS 3.6.D.3 will be revised to clarify ar>

tate the shutdown requirements when TS LCO 3.6.0.1 or 3.6,0.2 is not v, lied with.

These changes to TS Section 3.6.D. are acceptable.

The licensee proposes changes to SRs 4.6,0.1, 4.6.D.2, 4.6,0.3, and 4.6,0.4 for the safety and relief valves.

The proposed SR 4.6.D.1 permits that all safety and relief valves be tested, reinstalled or replaced with pretested spares once every 40 months.

The existing SR 4.6.D.1 requires that all valves be tested every two cycles.

This proposed change to allow the 40-month period to replace the existing two cycle test period requirement-is consistent with the Standard TS and is, therefore, acceptable to the staff.

The proposed SR 4.6.D.1 also requires that replacement spare valves be tested withi-the previous 40 months.

This meets the minimum requirements of the ASME Code regarding testing frequency and is, therefore, acceptable. The proposed requirement to test at least one safety valve and three relief valves each cycle remains unchanged, as does the table of safety valve setpoints.

There were several other minor wording changes proposed to these SRs to improve

. clarity and consistency with the Standard TS.

These changes are only editorial in nature and are, therefore, acceptable.

2.6 TS 3/4.6.E - Jet Pumps Existing TS LCO 3.6.E.1 will be revised to refer to defined MODES of operation.

LC0 3.6.E.1 also contains a statement that, if a specific surveillance cannot be met, an additional surveillance is to be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

This proposed amendment relocates this information in its entirety to proposed SR 4.6.E.1.c.

Existing TS LCO 3.6.E.1.a and 3.6.E.1.b will be revised and renumbered to proposed LCO 3.6.E.1.a, 3.6.E.1.a.1, and 3.6.E.1.a.2.

These proposed changes are being made to provide clarity within the LCO.

l A shutdown requirement has been proposed for TS LCO 3.6.E.1.a.2 to be consistent with the guidance provided by the Standard TS and to eliminate unnecessarily cycling the plant to the COLD SHUT 00WN condition as currently required in the DAEC TS.

j The staff finds these changes to TS Section 3/4.6.E acceptable.

2.7 TS 3/4.6 F - Jet Pumo Flow Mismatch Existing TS LC0 3.6.F.1 will be divided into two itemized sections proposed as TS LCOs 3.6.F.1 and 3.6.F.2.

Hinor editorial changes will be made to each LCO, in order to allow each to stand alone. These minor changes do not change the intent or requirements of the existing LCO.

Proposed TS LC0 3.6.F.3 and 3.6.F.3.a will be added as clarification and for consistency with the guidance provided by the Standard TS.

The addition of this LC0 allows two hours for the recirculation pump speeds to be restored within the above limits.

The current TS does not allow any time to restore the system to within limits before taking further action.

Existing TS LC0 3.6.F.2 will be revised and renumbered as LCO 3.6.F.3.b.

Existing SR 4.6.F.1 will be editorially revised to provide additional clarification and consistency replacing the words " checked and logged" with

" verified."

l Existing SR 4.6.F.2 will be editorially revised by changing the word

" Specification" to " Surveillance Requirement." The number referenced is a Surveillance Requirement number and is identified accordingly.

The staff finds these changes in TS Section 3/4.6.F acceptable.

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. 2.8 TS 3/4.6.G - Structural Intearity The licensee will combine the surveillance requirements for Inservice Inspection of ASME Code Class 1, 2, and 3 components (as required by existing SR 4.6.G.I.) and Inservice Testing requirements of ASME Code Class 1, 2, and 3 pumps and valves (as required by existing SR 4.6.G.2.) into proposed SR 4.6.G.I.

The l'icensee has also proposed that the reference of the 2nd Ten Year Inservice Inspection interval and the 2nd Ten Year Inservice Testing interval (as referenced in SRs 4.6 G.I.a. and 4.6.G.2.a.) be removed. These changes do not affect safety and are acceptable to the staff.

The licensee has proposed revising the existing LC0 3.6.G., " Structural Integrity," into LC0 3.6.G.I.

The existing LCO requires the structural integrity of the pressure boundaries be maintained in accordance with levels required by the original acceptance standard throughout the life of the plant.

LC0 3.6.G.I. now requires the structural integrity of ASME Section XI Code Class 1, 2, and 3 components be maintained in accordance with Inservice Inspection (for Code Class 1, 2, and 3 components) and Inservice Testing requirements (for Code Class 1, 2, and 3 pumps and valves) of Section XI of the ASME Boiler and Pressure Vessel Code, as required to be performed by SR 4.6.G.I.

The proposed LC0 is more specific than the existing requirement and is acceptable to the staff.

The licensee has proposed adding LC0 3.6.G.2., which will require that, under circumstances when the structural integrity of a ASME Code Class 1 or 2 component (s) does not conform to the requirements of the ASME Code Section XI, the structural integrity of the ASME Code Class 1 or 2 component (s) be restored within acceptable limits, or else that the affected component (s) be isolated prior to increasing the Reactor Coolant System (RCS) temperature above 212 F.

The licensee's basis for the addition was that it would make the requirements of TS Section 3.6.G. consistent with the corresponding requirements found in the Hope Creek Standard TS (NUREG-1202).

The staff informed the licensee, on March 10, 1994, that the RCS temperature, referred to for ASME Code Class I components in the corresponding Hope Creek Standard TS Action Statement, was 50 F + Nil Ductility Temperature (NDT, in

• F), not 212 F.

The licensee informed the staff, by conference call on March 24, 1994, that the standard operational practice during STARTUP at DAEC is to heat up the reactor prior to reaching (i.e., at a temperature very close to) 212 F, with subsequent closing of the reactor head vent.

This practice is delineated in a plant specific operating procedure.

This practice will keep the licensee to the right of Curve C in TS Figure 3.6-1 during startup, and thus prevent the licensee from pressurizing the reactor during startup above the Curve C limiting pressure (i.e., starting up in the safe operating regime). The staff have therefore, concluded that the 212 F temperature reference in the Class I component Action Statement is acceptable.

The staff concludes that LC0 3.6.G.2., as applied to ASME Code Class 2 i

components, is consistent with the intent of the applicable Action Statement

)

for ASME Code Class 2 components in NUREG-1202 (H0PE CREEK STANDARD TS).

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't

. The licensee has proposed adding LC0 3.6.G.3., which will require that, under circumstances when the structural integrity of a ASME Code Class 3 component (s) does not conform to the requirements of the ASME Code Section XI, the structural integrity of ASME Code Class 3 component (s) be restored within acceptable limits, or else that the affected component (s) be isolated from service.

LC0 3.6.G.3. is consistent with the corresponding Action Statement for ASME Code Class 3 components in TS Section 3.6.G. of the Hope Creek Standard TS (NUREG-1202). Therefore, the addition of proposed LC0 3.6.G.3. is acceptable to the staff.

The licensee has also proposed adding LC0 3.6.G.4., which would require that, with the reactor in the RUN, STARTUP, or HOT SHUTDOWN Modes of operation, should the requirements of proposed LC0 3.6.G.2. or 3.6.G.3. not be met, that the licensee perform an engineering evaluation to determine the effects of the component-(s) condition for continued operation and determine that the component (s) remain acceptable for continued operation, or else isolate the affected component (s) and follow the applicable system LCO.

The licensee's basis for the addition was that it would make the requirements of TS Section 3.6.G. consistent with the corresponding requirements found in the Hope Creek Standard TS (NUREG-1202).

The staff informed the 14:ensee, during a conference call on February 7,1994, that proposed LC0 3.6.G.4. was not consistent with the corresponding

" Structural Integrity" Section in the Hope Creek Standard IS, and that the proposed addition could potentially conflict with the LC0 Action Statements in Section 3.6.C., " Coolant Leakage," of the DAEC TS.

The licensee informed the staff, during the conversation of February 7, 1994, that it would withdraw the request for the additional section, LCO 3.6.G.4.

The licensee confirmed this by sending a revised LC0 3.6.G.4 to the staff on February 14, 1994.

2.9 TS 3/4.6.H - Shock Suppressors The licensee proposed changes to TS LCOs 3.6.H.1 and 3.6.H.2 to provide clarity by spelling-out specific operation Modes and to correct the specification section number referenced for consistency throughout the revised TS section.

Based on our review, we find that these changes are editorial in nature and are, therefore, acceptable.

The licensee proposed changes to TS SR 4.6.H.1 by replacing the existing visual examination schedule with an alternate visual examination schedule provided in Appendix B to Generic letter 90-09, " Alternative Requirements for Snubber Visual Inspection Intervals and Corrective Actions," dated December 11, 1990.

The existing TS schedule for snubber visual examination is based on the number of inoperable snubbers found during the previous visual examination irrespective of the total population of snubbers.

The proposed visual examination interval is based on the number of inoperable snubbers

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found during the previous visual examination in proportion to the size of the snubber population for each type of category.

The purpose of the alternate inspection schedule is to allow the licensee to perform visual examination and corrective actions without reducing confidence level provided by the existing visual examination schedule.

The existing inspection interval is based on a fuel cycle of 18 months. The proposed inspection interval is based on a fuel cycle of up to 24 months, and may be extended to as long as 48 months, depending on the number of unacceptable snubbers found during the previous visual examination.

This change is consistent with Generic Letter 90-09 and is, therefore, acceptable to the staff.

The licensee proposed that TS SR 4.6.H.2 regarding visual inspection acceptance criteria be revised to include a requirement that a review and evaluation be performed and documented to justify continued operation with an unacceptable snubber.

This proposed change is consistent with the requirement of the American Society of Mechanical Engineers (ASME) Operation and Maintenance Code (OM Code) regarding supported systems or components of which an unacceptable snubber is a part, and is, therefore, acceptable.

The licensee proposed to add SR 4.6.H.3, " Transient Event Inspection."

Existing SRs 4.6.H.3, 4.6.H.4 and 4.6.H.5 will be renumbered to SRs 4.6.H.4, 4.6.H.5 and 4.6.H.6, respectively.

The proposed SR 4.6.H.3 requires that in the event of a potential damaging transient, an inspection of all affected snubbers be performed and freedom-of-motion of affected mechanical snubbers be verifled within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for accessible systems and 6 months for inaccessible systems following such an event.

This proposed change is consistent with the Standard TS and with the ASME OM Code requirement regarding transient dynamic events (e.g., water hammer, steam hammers) that may affect snubber operability and is, therefore, acceptable. The other changes in the sections mentioned above are editorial in nature and are, therefore, acceptable.

The licensee proposed to add SRs 4.6.H.7 and 4.6.H.8 and to delete the existing SR 4.6.H.6 for the snubber service life monitoring.

The proposed SR 4.6.H.7 requires that replacement and repaired snubbers be tested to meet the 1

functional test criteria before installation in the unit.

This proposed change is consistent with the ASME OM Code snubber inservice inspection requirements regarding the replacement and modification of snubbers and is, therefore, acceptable. With regard to the service life monitoring, both the existing SR 4.6.H.6 and the proposed SR 4.6.H.8 require that the service life of all snubbers be monitored, and that documentation be maintained to provide assurance that the service life of each snubber is not exceeded between i

surveillance inspection.

In addition, the existing SR 4.6.H.6 requires that in the event that the indicated service life will be exceeded, prior to the next scheduled inspection, the snubber be reevaluated, replaced or reconditioned to extend its service life beyond the next scheduled inspection.

l The proposed SR 4.6.H.8 requires that all critical parts of the snubber be j

monitored and replaced as necessary, so that the snubber service life will not be exceeded during a period when the snubber is required to be operable. This proposed SR 4.6.H.8 appears to have incorporated the intent of the existing SR 4.6.H.6 regarding the snubber service life monitoring and is, therefore, acceptable.

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• . 2.10 Bases Sections 3/4.6.A-H The Bases Sections for 3/4.6. A-E, G & H have been revised to reflect the proposed changes and are acceptable.

Bases Section 3/4.6.F was not changed.

2.11 Supolemental Information i

The supplemental information provided on February 14, 1994, did not change the proposed no significant hazards determination.

3.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Iowa State official was notified of the proposed issuance of the amendment. The State official had no comments.

4.0 ENVIRONMENTAL CONSIDERATION

S This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements.

The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluent that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public con. ment on such finding (58 FR 39052).

Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

5.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors:

J. Medoff l

W. LaFave G. Hammer C. Wu R. Frahm R. Pulsifer l

J. Minns Date: November 17, 1994 l

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