ML20081B240

Safety Evaluation Supporting Amends 204 & 182 to Licenses DPR-53 & DPR-69,respectively

ML20081B240
Person / Time
Site:	Calvert Cliffs
Issue date:	03/14/1995
From:	Office of Nuclear Reactor Regulation
To:
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ML20081B239	List: ML20081B240 ML20081B247
References
NUDOCS 9503160100
Download: ML20081B240 (8)

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UNITED STATES

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. NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20$56 0001 o

49.....,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 204 TO FACILITY OPERATING LICENSE NO. DPR-53 AND AMENDMENT NO.182 TO FACILITY OPERATING LICENSE NO. DPR-69 BALTIMORE GAS AND ELECTRIC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT. UNIT NOS. 1 AND 2 DOCKET NOS. 50-317 AND 50-318 I

1.0 INTRODUCTION

By letter dated December 8,1993, as supplemented on March 2,1995, the Baltimore Gas and Electric Company (BG&E) submitted a request for changes to the Calvert Cliffs Nuclear Power Plant, Units Nos. I and 2 (CC1/2) Technical Specifications (TSs). The requested changes would, for the most part, adopti the improved Standard Technical Specifications (STS) format and content of Section 5.0, " Design Features," for the CC1/2 TSs, as modified by approved changes to the improved STSs.

In addition, BG&E initially proposed to eliminate two subsections of the improved STSs. The March 2, 1995, letter provided clarifying information and withdrew the request to delete two subsections. These revisions did not change the initial proposed no significant hazards consideration determination.

2.0 BACKGROUND

Section 182a of the Atomic Energy Act (the "Act") requires applicants for nuclear power plant operating licenses to state TSs to be included as part of the license. The Commission's regulatory requirements related to the content of TSs are set forth in 10 CFR f 50.36. That regulation requires that the TSs include items in five specific categories, including:

(1) safety limits, limiting safety system settings and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls.

However, the regulation does not specify the particular requirements to be included in a plant's TS.

The Commission has provided guidance for the contents of TSs in its " Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" (" Final Policy Statement"), 58 Fed. Reg. 39132 (July 22,1993), in which the Commission indicated that compliance with the Final Policy Statement

• isfies 5 182a of the Act.

In particular, the Commission indicated that cutain items could be relocated from the TSs to licensee-controlled documents, consistent with the standard enunciated in Portland General Electric Co. (Trojan Nuclear Plant), ALAB-531, 9 NRC 263, 273 (1979).

In that case, the Atomic Safety and Licensing Appeal Board indicated that " technical specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed necessary 9503160100 950314 PDR ADOCK 05000317 p

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9 to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety." The policy statement encouraged licensees to adopt the applicable improved STSs and provided some guidance for the conversion from the present plant specific TSs to the improved Standard TSs. However, specific guidance for converting the design features section of TSs was not provided in the policy statement. CC1/2 are Combustion Engineering (CE) designed units and the improved STSs for CE plants was published as NUREG-1432 in September 1992.

3.0 DISCUSSION While the example design features section in NUREG *.432 provides a standard for the design features of a typical CE plant, as noted above, there is no specific guidance for a licenree to follow when updating the design features section of its plant specific TSs.

In this case, BGEE conducted a review of the existing legislation and regulations potentially impacting the content of the design features section of TSs for guidance in updating that portion of its current TSs.

Three applicable references were identified.

Section 182.a of the Atomic Energy Act of 1954, as amended (the Act) " License Applications," states, in part:

In connection with applications for licenses to operate production or utilization facilities, the applicant shall state such technical specifications, including information of the amount, kind and source of special nuclear materials required, the place of the use, the specific characteristics of the facility, and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that utilization or production of special nuclear material will be in accord with common defense and security of the public.

Such technical specifications shall be a part of any license issued.

j 10 CFR 50.36(c)(4) states, in part, " Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1),

(2), (3) of this section [ Safety Limits, Limiting Conditions for Operations, and Surveillance Requirements)." Section 182.a of the Act does not specify the design features that must be included in the TSs.

10 CFR 50.59, " Changes, tests and experiments," provides the criteria for determining if a change to the features of a facility, as described in its Final Safety Analysis Report (FSAR), is an unreviewed safety question. The i

criteria in 10 CFR 50.59 for making this determination are:

(1) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; (2) if a possibility for an accident or malfunct00n of a different type than any evaluated previously in the safety analysis report may be created; or (3) if the margin of safety as defined in the basis for any TS is reduced.

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~ 1 BG&E indicated that those features of CC1/2 for which any change would have an

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immediate and significant impact on safety would not be appropriate for being controlled using 10 CFR 50.59. This is consistent with 10 CFR 50.36(c)(4).

c With the above citations as references, BG&E concluded that the following criteria are appropriate to use in determining what information should be.

l placed in the design features section of the CC1/2 TSs:

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The amount, kind, and source of special nuclear material required;'

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The place of the use of the special nuclear material; and 3.

Those features of the facility, such as na w<ials of construction and l

geometric arrangements, which it mered o' mWified would have an immediate and significant effect on safety and are not covered in the i

safety limits, limiting conditions' ter opmtim (LCO), or surveillance j

requirements of the TSs.

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Based on the above criteria, BG&E proposes to change Section 5.0, " Design j

Features," of the CC1/2 TSs by eliminating some subsections, including some_

contained in NUREG-1432, and modifying others to conform to the examples

.provided in NUREG-1432. The subsections which BG&E proposes to change or modify are; 5.1 Site Location (including Figures ~ 5.1-1 and 5.1-2), 5.3 Reactor Core, and 5.6.1 and 5.6.2 of 5.6 Fuel Storage. The subsections proposed for elimination are; 5.2. Containment, 5.3.3 Control Element Assemblies, 5.4 l

Reactor Coolant, 5.5 Meteorological Tower Location, 5.6.3 Drainage and 5.7 l

r Component Cyclic or Transients (including Table 5.7-1).

1 4.0 EVALUATION The NRC staff has reviewed BG&E's proposed criteria as well as the background l

information that provided the bases for the criteria. With one exception, i

relating to the immediate effect on safety, the NRC staff agrees with the

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criteria chosen by the BG&E. An immediate effect on safety is not stated or 1

inferred in the Act or 50.36(c)(4) as a necessary condition for inclusion of a matter in the TSs. A significant effect on safety, whether immediate or delayed, must be considered. Thus, the NRC staff's evaluation that follows did not consider the timing when assessing the safety significance of BG&E's proposed changes.

The NRC staff reviewed 2G&E's proposed changes individually against both the t

proposed criteria, with the noted exception, and the examples provided in the design features section of NUREG-1432 as follows:

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a.

5.1 Site Location (including Figures 5.1-1 and 5.1-2) - Section 182.a of l

the Act requires that the place of use of the special nuclear material be specified.

Presently, subsection 5.1.1 meets that requirement by reference to Figure 5.1-1 which is a map of the site.

In BG&E's proposed 1

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Section 5.1,.the place of use is specified by a description rather than a

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. map. This is consistent wid NUREG-1432 which indicates that the site l

shall be described g.t as.shown in figures.

l As-noted, BG&E >roposes to eliminate the present maps showing the site and.

exclusion area >oundaries ( Figure 5.1-1) and also the Low Population Zone

.(Figure 5.1-2) and relocate them to Chapter 2 of the Updated Final Safety-Analysis Report (UFSAR). BG&E notes that this is consistent with the Standard Review Plan, NUREG-0800,. which indicates that the site details are to be i

provided in Chapter 2 of the FSAR. (Although the example.in NUREG-1432 l

indicates the use.of maps for defining the site boundaries and low population zone, a description may provide an equivalent presentation of. the existing TS

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intent as long as a minimum exclusion area boundary is included'in the j

description. By letter dated March 2,1995, BG&E included this clarification j

in the proposed description. Thus, in accordance with 10 CFR Part-100, the-i site description includes a minimum distance to the Exclusion Area Boundary to 1

ensure that the area, for which the licensee has the authority to determine l

all activities including the-exclusion or removal of personnel; and property.

1 from the area, is clear'y associated with the " place of use" referred to in i

Section 182.a of the Act. The inclusion of the maps in the UFSAR will ensure.

that any change to either the boundaries or~the zone will have to be evaluated using the 10 CFR 50.5g process.

Based on the above discussion, the NRC staff finds this proposed change to.

Section 5.1, including the description of the Exclusion Area Boundary,-

acceptable.

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b.

5.2 Containment - BG&E proposes that this section be completely-i eliminated. However, certain modifications or alterations to the containment would have a significant impact on plant safety and therefore j

they are required to be controlled by TSs. BG&E notes that accounting for such changes is already adequately controlled by the containment limiting conditions for operations in Section 3/4.6 of the CC1/2 TSs and need not-i be specified in the design features section.

Further, the information

.i contained in Section 5.2 of the TSs, which is to be eliminated'by this proposed change, is presently in UFSAR sections,-such as'5.1.2.1 and 5.1.5.1.

Any change to the UFSAR, as previously noted, would have to be evaluated using the 10 CFR 50.59 process prior to being implemented.

Finally, the example provided in NUREG-1432 does not contain this information.

Based on the above discussion, the NRC staff finds this proposed change to eliminate the existing Section 5.2 acceptable.

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5.3 Reactor Core - Recognizing that the Act requires the amount, type and source of special nuclear material be specified in the TSs, BG&E proposes to modify the information of Subsections 5.3.1 and 5.3.2 into a'new Section 5.2 consistent with the example provided in NUREG-1432 including i

the removal of the fuel enrichment information. BG&E further proposes to j

eliminate Subsection 5.3.3 which is not consistent with the NUREG.

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. The fuel enrichment information indirectly quantifies the amount of special nuclear material in use.

Consistent with the Act that information must be controlled in the TSs. The removal of the enrichment information from the design features section of TSs is not inconsistent with the requirements of the Act in that the various CC1/2 TSs controlling reactor core reactivit/

adequately account for the fuel enrichment limits. The NRC staff identified one clarification needed for the existing and proposed descriptions of the fuel assemblies. The existing and proposed descriptions refer to the fuel assemblies as zirconium alloy fuel rods.

10 CFR 50.46 specifically states cylindrical zircaloy or zirlo cladding. Although the existing zirconium alloy fuel rods comply with 10 CFR 50.46, by letter dated March 2, 1995, BG&E revised the wording to be the same as 10 CFR 50.46 to assure that any future changes will also be in compliance with that regulation. Therefore, the portion of the proposed Section 5.2 which replaces Subsections 5.3.1 and 5.3.2 contains the necessary information required by the evaluation criteria and is consistent with the example provided in NUREG-1432.

Eliminating Subsection 5.3.3, " Control Element Assemblies," completely from design features section, as noted, is inconsistent with the example provided in NUREG-1432. BG&E stated in its initial submittal that the safety significant aspects of the control element assemblies (CEAs) which require inclusion in the TSs (reactivity worth and insertion times) are adequately controlled elsewhere in the TSs and need not be included in the design features section.

Further, BG&E maintained that the mechanical design of the CEAs is specified in the UFSAR and that the less critical aspects of the CEAs performance can be adequately controlled by using the 10 CFR 50.59 process.

The NRC staff does not agree with the elimination of Subsection 5.3.3.

Should the number of CEAs or their materials of construction be altered they could have a significant impact on safety and therefore must be controlled by TSs and not by the 10 CFR 50.59 process. The maximum overall core reactivity worth, not specifically the miniinum CEA reactivity worth, is adequately controlled by the other portions of the TSs. The other critical CEA TSs (e.g., insertion times and insertion limits) are predicated on there being a specific number of assemblies and that the assemblies be constructed of specified materials. Therefore, the NRC staff determined that the TSs proposed by BG&E are not adequat? in assuring the nuclear performance of the CEAs and the information in the current design features description of the CEAs should be retained.

By letter dated March 2, 1995, BG&E modified its initial request to delete this proposed subsection indicating the existing information would be retained.

Based on the above discussion, the NRC staff finds this proposed change, including the inclusion of cylindrical zircaloy or zirlo cladding in the fuel assemblies description, for a new Sc9. ion 5.2 acceptable. The new Section 5.2 will include Subsection 5.2.1 which will incorporate the information in the old Subsections 5.3.1 and 5.3.2, as detailed above, and a new Subsection 5.2.2 4

which will retain the information in the old Subsection 5.3.3.

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5.4: Reactor. Coolant System - BG&E proposes to eliminate this section.;

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.BG&E has determined that other CC1/2 TSs adequately contro11the' reactor:

.i coolant. system parameters, such.as; temperature, pressure, and boundary l

degradation, which could have.a significant' impact on safety. 'In l

-addition, BG&E states that the specific information eliminated by the8

-i removal of Subsection 5.4 is already contained in the UFSAR Section 4.1.1, i

and Table 4-1 and would be' adequately controlled by the 10 CFR 50.59 l

process.

l The NRC staff has determined that this proposed' changezis cons' stent with the N

i requirements of the Act and the regulations cited earlier as well as the -

example design festures section of NUREG-1432. Therefore, based on the above.

1 discussion, the NRC staff finds this proposed change to eliminate Section 5.4

-l acceptable.

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5.5 Meteorological Tower Location - BG&E proposes to eliminate this.

l section which specifies the tower's location by a reference to Figure 5.1-1.

BG&E has determined that the tower's location does not meet the i

criteria for inclusion in the'TSs. Further, BG&E states that the inclusion of Figure 5.1-1 into the UFSAR will ensure that any change to-l the tower's location is adequately evaluated by using the 10 CFR 50.5g l

process.

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The NRC staff has determined that the proposed elimination of this section and -

inclusion of the information into the UFSAR is appropriate in that.the information contained in this section is not required to be controlled by the.

TSs.

Further, the NRC staff finds that the' proposed change ~is consistent with NUREG-1432. Therefore, based on the above discussion, the NRC staff finds this proposed change to eliminate Section 5.5' acceptable.-

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5.6. Fuel Storage - BG&E initially proposed to:

(1) retain the

.l information in present subsections 5.6.1 and 5.6.2 in.the format of.

Subsection 4.3.1 of the example provided in NUREG-1432;-(2) retain the.

l information in present subsection 5.6.4 and; (3) eliminate the information of Subsection 5.6.3 Drainage.

i The NRC staff has determined that the proposed modifications of the information presently contained in Subsections 5.6.1 and 5.6.2, which includes i

incorporation of that information into a new Subsection 5.3.1, along with e

relocation of present Subsection 5.G.4 into a new Subsection 5.3.3 is appropriate. The proposed changes do not eliminate any existing infornit

,n required to be controlled by TSs and the proposed changes are consistent with i

NUREG-1432.

t With respect to elimination of Subsection 5.6.3, the design considerations required to prevent inadvertent draining of the spent ~ fuel pool (SFP), the staff finds this initial proposal unacceptable. BG&E's rationale for eliminating this subsection is that it does not meet the evaluation criteria.

1 BG&E indicates that it does not describe geometry or materials of construction and the requirements are contained in other portions of the TSs.

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' BG&E's first point concerning the design considerations required to be considered is incorrect because it is based on a limited interpretation of 10 CFR 50.36(c)(4). Rather than restricting applicability to the design considerations of geometry and materials of construction, the regulation simply states those as examples. All design considerations that would have a-significant impact on safety and that are not accounted for in other TSs must be included in therdesign features section. Clearly, the inadvertent draining of the SFP would have a significant impact (though it may not be immediate) and therefore must be considered in the design features section if inadvertent draining of the SFP is not accounted for elsewhere in TSs.

BG&E indicates that the inadvertent draining of the SFP is already accounted for in TS 3.9.11 (not TS 3.9.1.1 as incorrectly stated in the BG&E's initial submittal). The LC0 for TS 3.9.11 would detect a decrease in level due to an inadvertent draining. However, detection of inadvertent draining is not the same as preventing it. TS 3.9.11 is written (as explained in the TS Bases) only to ensure that an adequate amount of water covers the spent fuel elements in the event of a fuel handling accident.

If there is not enough water covering the elements, the TS requires that the handling of fuel in the SFP and the transfer of loads over it are to be suspended. Once those actions are i

taken, the TS requirements have been met. The TS does further state that the SFP level should be returned to the specified level within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. However, l

that requirement (which includes no guidance on how to proceed if the l

specified level cannot be achieved within that time period) presupposes all l

drain paths are known and that normal makeup will be adequate to restore level. The control of drain paths and the adequacy of normal makeup are only assured by the Design Features Subsection 5.3 which states, "The spent fuel storage pool is designed and maintained to prevent inadvertent draining of the pool below elevation 63 feet." Thus, the NRC staff determined that deleting Subsection 5.3.3 is inappropriate. By letter dated March 2, 1995, BG&E

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modified its initial request to delete this subsection indicating the existing information would be retained.

i Based on the above discussion, the NRC staff concludes that this proposed change for a new Section 5.3 is acceptable. The existing information in old Subsection 5.6.3 shall be retained as a new Subsection 5.3.2.

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5.7 Component Cyclic or Transient Limits (including Table 5.7-1) - BG&E proposes to eliminate the information contained in this section.

BG&E has determined that the information does not meet the criteria for inclusion, maintains that the presence of the information in the UFSAR (Section 4.1.1 and 4.1.3.2) will ensure that these considerations will be adequately controlled by the 10 CFR 50.59 process, and notes that this proposed change is consistent with the example section of NUREG-1432.

In addition, the component cyclic or transient limits need not be retained in the design features section because adequate controls for these requirements exist in 10 CFR 50.55a, as well as 10 CFR 50.59. The NRC staff agrees with this determination.

r-Q Based on the above discussion, the NRC staff concludes that this proposed change to eliminate Section 5.7 is acceptable.

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In summary, the proposed changes to the Design Features section of the Calvert Cliffs 1/2 TSs, as detailed in this Safety Evaluation,. satisfy Section 182.a of the Act and 10 CFR 50.36(c)(4). Further, they are not required to_ obviate i

the possibility of an abnormal situation or event giving rise to an immediate y

threat to the public health and safety.

In addition, the NRC staff finds that sufficient regulatory controls exist under 10 CFR 50.59 to ensure that future changes to these items are acceptable.

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5.0 STATE CONSULTATION

1 In accordance with the Commission's regulations, the Maryland State official l

.was notified of the proposed issuance of the amendments. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

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The amendments change a requirement with respect to installation or..use of a facility component located within the restricted area as defined in 10 CFR-Part 20. The NRC staff has determined that the amendments involve no.

significant increase in the amounts, and no significant change in the types,

i of any effluents 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding (59 FR 2861). Accordingly, the amendments meet 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 i

environmental assessment need be prepared in connection with the issuance of the amendments.

7.0 CONCLUSION

- The Commission 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 ranner, (2) such activities will be conducted in compliance with the Commission's regulations, 1

and (3) the issuance of the amendments will not be inimical to the common j

defense and security or to the health and safety of the public.

l Principal Contributors:

J. Luehman r

D. Mcdonald I

i Date:

March 14, 1995

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