ML20140G541
| ML20140G541 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 06/12/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20140G530 | List: |
| References | |
| NUDOCS 9706160335 | |
| Download: ML20140G541 (13) | |
Text
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-, a UNITED STATES l
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20656 4001
%...../
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
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RELATED TO AMENDMENT NO. 160 TO FACILITY OPERATING LICENSE NO. DPR-19 E
AND AMENDMENT NO. 155 TO FACILITY OPERATING LICENSE NO. DPR-25 COMMONWEALTH EDISON COMPANY DRESDEN NUCLEAR POWER STATION. UNITS 2 AND 3 3
]
DOCKET NOS. 50-237 AND 50-249
1.0 INTRODUCTION
By letter dated June 20, 1996, as supplemented by letters dated December 30, 1996, and March 5, 1997, Commonwealth Edison Company (Comed, the licensee),
requested changes to the Dresden Nuclear Power Station, Units 2 and 3, Technical Specificatins (TS). Dresden, Units 2 and 3, currently use Siemens Power Corporation (SPC) fuel and licen:kg methodologies that were previously rt: viewed and approved by the NRC and are listed in TS 6.9.6.b..
Comed renewed the SPC contrae.t to supply fuel and sui. port services for Dresden, Unit 3, Cycle 15, and Dresden, Unit 2, Cycle 16. Two significant changes under the
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new SPC contract include introduction of an advanced fuel design, ATRIUM-98, and the use of a previously approved Siemens' loss-of-coolant accident (LOCA) methodology.
The use of ATRIUM-9B has been reviewed and approved by the staff, for full power operation at Quad Cities Unit 2.
Thus, the proposed changes to the Dresden, Units 2 and 3, TS would incorporate the above
- L methodology in the list of approved methodologies used in establishing the cycle-specific thermal limits. Other minor editorial changes are also 7
- proposed, s
In March 1997, the NRC staff performed an audit of the application of Advanced Nuclear Fuel for Boiling Water Reactors (ANFB) methods to ATRIUM-9B fuel. The s
staff raised concerns associated with the ATRIUM-9B fuel additive consi. ant uncertainty used as an input parameter to the NRC-approved methodology that is used for the cdculation of Minimum Critical Power Ratio (MCPR).
In response to the audit finlings, SPC submitted a generic topical report, ANF-ll25(P),
Supplement 1, Appendix 0, dated April 18, 1997, which is currently under staff review for the future reload analysis in the safety limit MCPR calculation.
The staff schedule for the review of the topical report will not be timely enough for the resolution of the I.TRDM-9B MCPR issue to support reload and restart of Dresden, Unit 3. By letters dated May 2, 1997, May 6, 1997, and June 10, 1997, the licensee provided additional information concerning the MCPR issues and how it will affect the Dresden, Unit 3, fuel cycle D3R15 and b*
g provided additicnal information concerning the ATRIUM-98 fuel design and g
shutdown margin which are applicable during refueling ar.d shutdown.
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I To be more timely and support the reload schedule for Dresden, Unit 3, the staff chose to issue the amendment in two parts. On May 16, 1997, the NRC issued Amendment Nos. 159 and 154. The amendments modified Tsction 5.3.A,
" Design Features," of the TS to reflect the ATRIUM-9B fuel design end added two SPC topical reports in TS Section 6.9.A.6, " Core Operating Limits Report."
The change allowed ATRIUM-9B fuel to be loaded into the core only under Operational Modes 3 (Hot Shutdown), 4 (Cold Shutdown), and 5 (Refueling) and l
does not permit startup or power operation using the ATRIUM-9B fuel.
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j This amendment addresses, on a cycle-specific basis, th; autstanding issues s
concerning the uncertainty of ANFB additive constants used for 9X9 fuels with an internal water channel in the MCPR safety limit analysis. This is discussed in detail in Section 2.9 of this Safety Evaluation (SE). A similar cycle-specific amendment has been approved for Quad Cities.
In addition, this amendment addresses all other requested changes to the TS.
By letters dated May 2, 1997, May 6, 1997, and June 10, 1997, Comed provided additional information. These letters provided additional information that did not change the initial proposed no significant hazards consideration determination.
2.0 EVALUATION 2.1 Mechanical Desian The ATRIUM-9B fuel design is a 9xb 1r.tice design with an internal water box to enhance neutron moderation. The ATRIUM-9B fuel mechanical design was analyzed and assessed by Siemens according to the approved methodology entitled, " Generic Mechanical Design for Advanced Nuclear Fuels 9x9-IX and 9x9-9X BWR Reload Fuel," ANF-89-014(P)(A), Revision 1, Supplement 1.
This methodology was previously reviewed and approved by the NRC in 1991 and is listed as a reference in Section 6.9.6.b of the Dresden TS. The ATRIUM-9B fuel mechanical design followed the approved methodology and, therefore, is acceptable for Dresden Nuclear Power Station, Unit 3, Cycic 15.
2.2 Safety limits and Limitino Safety System Settinas 2.2.1 Section 2.1.D - Reactor Vessel Water Level The licensee has proposed to remove the word "the" prior to " active" and add a footnote following the word " fuel." The proposed change would be as follows:
2.1.D The reactor vessel water level shall be greater than or equal to twel/e inches above the top of active irradiated fuel").
The footnote would state:
(a) The top of active irradiated fuel is defined to be 360 inches above vessel zero.
1
. The licensee stated that current fuel designs, including ATRIUM-9B, incorporate slight design variations in the length of the active fuel and, therefore, the true location of the top of active (irradiated) fuel may be difficult to distinguish. However, this fixed reactor vessel reference point, i.e., 360 inches above vessel zero, for top of active fuel is used for -the cutomatic initiations associated with both accident and transient analyses.
This reference point can also be found in other TS such as the Emergency Core Cooling System (ECCS) and Isolation Instrumentation Tables.
Based on this information, the staff finds the inclusion of the footnote and the editorial change acceptable.
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2.2.2 Table 2.2. A Reactor Protection System Instrumentation Setooints The licensee also proposed to add the above footnote to Table 2.2. A-1, Reactor Protection System Instrumentation Setpoints. The proposed change would be as follows:
1 Functional Unit Trio setooint j
- 4. Reactor Vessel Water Level Low 2144 inches above top of active fuel" This change causes footnote b on Table 2.2.A-1 to become footnote c.
Since footnote b is the same as footnote (a) in Section 2.1.D, the staff finds the incorporation of the footnote and the associated editorial change acceptable.
2.3 Safety Limits bases The licensee proposed an editorial change to the Section 2.1, third paragraph, of the Bases. The current wording states "The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an A00."
i The proposed change would state the sentence as follows:
The fuel cladding integrity limit is set such that no fuel damage is
'J!culated to occur as a result of an A00.
The staff concludes that the change clarifies the meaning of the sentence and is acceptable.
In Section 2.1.B, THERMAL POWER, High Pressure and High Flow - tiie licensee proposed editorial changes to paragraph one. The current wording of the last sentence states that "Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9 percent of the fuel rods in the core are expected to avoid boiling transition considering the power riistribution within the core and all uncertainties."
The editorial change would have the last two sentences of paragraph one consist of the following:
Therefore, the fuel cladding integrity Safety Limit is defined such that, with the limiting fuel assembly at the MCPR Safety Limit, more than 99.9% of the fuel rods in the core are expected to avoid boiling
_ _ _.. _.. _. _.. _ _ _ ~. _
-l transition. This includes consideration of the power distribution within the core and all uncertainties.
The staff notes that this wording is consistent with the bases in Improved Standard Technical Specifications, NUREG-1433, Revision 1, and, therefore, it is acceptable.
In section 2.1.D, Reactor Vessel Water Level, the licensee proposed the incorporation of a few words to clarify the last sentence of the paragraph.
i The current wording is "The top of active fuel is 360 inches above vessel zero." The licensee has proposed that the sentence be as follows:
The top of active irradiated fuel is defined to be 360 inches above vessel zero.
The staff finds this consistent with the above footnotes and acceptable.
2.4 Limitina Safety System Settinas Bases Section 2.2.A.1, Reactor Protection System Instrumentation Setpoints -
Intermediate Range Monitor, Neutron Flux - High - the licensee proposed an editorial change to the third paragraph. The sentence with the proposed change states that "The results of this analysis show that the reactor is scrammed and peak power is limited to 1% of rated power, thus, maintaining minimum critical power ratio (MCPR) above the fuel cladding integrity Safety Limit." The licensee proposed to change 1. percent to 7.7 percent to reflect i
the correct value in the updated final safety analysis report (UFSAR) and safety analysis report (SAR)' analysis.
Section 7.6.1.4.3 of the UFSAR cites, in graphical form, 7.7 percent as the power level at which the intermediate-range monitors (IRM) terminate the low power rod withdrawal error (RWE) transient. Based on this information, the staff finds this editorial change acceptable.
j In Section 2.2.A.4, "eactor Protection System Instrumentation Setpoints -
Reactor Vessel WRa Level - Low - the licensee proposed to add a clarification of tne top of active fuel at the end of the paragraph. The proposed last sentence of the paragraph would read "The top of active fuel is defined to be 360 inches above vessel zero." This statement is consistent with the footnotes and other sections of the bases and, therefore, is acceptable.
2.5 Instrumentation Bases The' licensee proposed a clarification to Section 3/4.2, Instrumentation. - The licensee proposed to add the following sentonces to the bottom of the paragraph.
Current fuel dest.s incorporate slight variations in the length of the active fuel and, thu; the actual top of active fuel, when compared to the original fuel designs. Safety Limits, water level instrument
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setpoints and associated LCOs refer to the top of active fuel.
In these cases, the top of active fuel is defined as 360 inches above vessel zero.
Licensing analyses, both accident and transient, utilize this l_
definition for the automatic initiations associated with these events.
The prepcsed additions provide a clear definition and use of the top of active t
fuel reference point. The staff finds this addition to the bases acceptable.
1 In section 3/4.2.A, Isolation Actuation Instrumentation, the l'icensee proposed the removal of the following sentences from the second paragraph since j
retrofit 8x8 fuel is no longer used at Dresden, Units 2 and 3.
Retrofit 8x8 fuel has an active fuel length 1.24 inches longer than earlier fuel designs. However, present trip setpoints were used in the j
loss-of-coolant accident (LOCA) analysis for Dresden, Units 2 and 3.
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The staff finds the removal of these sentences acceptable.
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2.6 Reactivity Control Bases In TS 3.3.8, Reactivity Anomalies, requires that the reactivity equivalence of 1
the difference between the actual critical control rod configuration and the j
predicted control rod configuration shall not exceed 1 percent Ak/k. This limit ensures that plant operation is maintained within the assumptions of the safety analyses.
The licensee proposed to include an alternative to i
monitoring reactivity anomalies in the TS bases. The SPC core monitoring code, Powerplex, provides the capability to monitor actual K,,, versus predicted K.,,.
This method is currently used at Dresden to monitor
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reactivity anomalies. Thus, the following will be added to Bases Section l
3/4.3.B, Reactivity Anomalies:
Alternatively, monitored K,,imulator code.can be compared with the predicted K,,, as j
calculated by the 3-D core s When the reactor core is critical or in normal power operation, a reactivity balance exists and the net reactivity is zero. A comparison of predicted and measured reactivity is convenient under such a balance, since parameters are i
being maintained relatively stable under steady state power conditions. The staff notes that this proposed change only revises the current method of j
measuring the difference between predicted and monitored core reactivity and i
does not change the required limit, therefore, the change is acceptable.
5 In sections 3/4.3.D, 3/4.3.E, and 3/4.3.F, Control Rod Maximum Scram Insertion Times, Control Rod Average Scram Insertion Times, and Four Control Rod Group Scram Insertion Time'., the licensee proposed to remove the following comments:
first paragraph:
"(as adjusted for statistical variation in the observed data);"
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second paragraph: "In the statistical treatment of the limiting transients, a statistical distribution of total scram delay is used rather than the bounding value described above;"
third paragraph:
" Observed plant data or Technical Specification limits were used to determine the average scram performance used in the transient analyses, and the results of each set of control rod scram tests performed during the current cycle are compared against earlier results to verify that the performe.nce of the control rod insertion system has not-changed significantly;" and fourth paragraph: "If test results should be determined to fall outside of 5
the statistical population defining the scram performance characteristics used in the transient' analyses, a re-determination of thermal margin requirements is undertaken as required by Specification 3.ll.C.
A smaller test sample than that required by these specifications is not statistically significant and should not be used in the re-determination of thermal margins."
l The licensee stated that the above information is based on past data which is a General Electric (GE) methodology. Current SPC methods used to evaluate the 5 percent, 20 percent, 50 percent and 90 percent control rod scram insertion times, collected during the performance of the scram timing Surveillance Requirement 4.3.D, will replace the above information as follows:
Transient analyses are performed for both Technical Specification Scram Speed (TSSS) and Nominal Scram Speed (NSS) insertion times. These analyses result in the establishment of the fuel cycle dependent TSSS MCPR operating limits and NSS MCPR operating limits which are presented in the core operating limit report (COLR).
Results of the control rod scram timing tests performed during the current fuel cycle are used to determine the operating limit for MCPR.
Following the completion of each set of scram time testing, the results will be compared with the assumptions used in the transient analysis to verify the applicability of the MCPR operating limits.
Prior to the initial scram time testing for an operating cycle, the MCPR operating limits will be based on the TSSS insertion times.
The NSS insertion times are typically faster than the TSSS insertion times, thus, the NSS insertion times are used to calculate the NSS MCPR operating limit.
If any of the average scram insertion times do not meet the NSS times, the TSSS MCPR operating limit is used. TS 3.11.C, Minimum Critical Power Ratio, requires that the MCPR shall be equal to or greater than the MCPR operating limit specified in the COLR. These changes to the bases clarify how Dresden, Units 2 and 3, will uso the NRC-approved SPC methodology, as listed in Section 6.9.6.b of the TS, and how it is used to meet TS 3.II.C.
Based on this information, tim changes to Section 3/4.3.D, 3.E, and 3.F Bases are acceptable.
.._ _ _._ 2.7 Primary System Boundary Bases In Sections 3/4.6.E and 3/4.6.F, the licensee proposed to add the following sentence to the middle of the first paragraph:
SPC methodology determines the most limiting pressurization transient.
each cycle.
The addition of this statement clarifies how Dresden will use the NRC-approved SPC methodology for analyzing the overpressurization event and, therefore, is acceptable.
2.8 Power Distribution limits Bases The licensee proposed changes to Sections 3/4.11.A, 3/4.11.B, and 3/4.11.C, Average Planar Linear Heat Generation Rate (APLHGR), Transient Linear Heat Generation Rate-(TLHGR), and Minimum Critical Power Ratio (MCPR), in order to provide clarification of the SPC methodologies listed in TS 6.9.6.b that were previously approved for the application of thermal limits. TS 3.11.A requires that all APLHGR for each type of fuel as a function of b:pdle average exposure shall not exceed the li: nits specified in the COLR. The licensee proposed to replace the first two paragraphs in Section 3/4.11.A with the following to describe the SPC methodology:
This specification assures that the peak cladding temperature following a postulated design basis loss-of-coolant accident will not exceed the Peak Cladding Temperature (PCT) and maximum oxidation limits specified in 10 CFR 50.46.
The calculational procedure used to establish the Average Planar Linear Heat Generation Rate (APLHGR) operating limits is based on a loss-of-coolant accident analysis. This analysis is performed using calculational models which are consistent with the requirements of Appendix K of 10 CFR Part 50.
The PCT following a postulated loss-of-coolant accident is primarily a function of the initial condition's average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod-to-rod power distribution within the assembly.
The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits for two-loop and single-loop operation are specified in the Core Operating Limits Report (COLR).
e staff finds the replacement of the first two paragraphs in Section
.i 3/4.11.A with the above paragraphs acceptable.
TS 3.11.B requires that the TLHGR shall be maintained such that the Fuel Design Limiting Ratio for Centerline Melt (FDLRC) is less than or equal to 1.0.
The licensee proposed to replace the last sentence in the first paragraph of Section 3/4.11.B with the following sentences:
r
_ _ _ _ The APRM scram settings must be adjusted to ensure that the LHGR transient limit (TLHGR) is not violated for any power distribution.
This is accomplished by using FDLRC. The APRM scram setting is decreased in accordance with the formula in Specification 3.11.B when FDLRC is greater than 1.0.
The adjustment may also be accomplished by increasing the gain of the APRM by FDLRC. This provides the same degree of protection as reducing the trip setting by 1/FDLRC by raising the initial APRM reading closer to the trip setting such that a scram would be received at the same point in a transient as if the trip setting had been reduced.
The second paragraph provides clarification of LC0 Action Statements 3.11.B.2 and 3.11.B.3.
Therefore, the above replacement paragraphs clarify the SPC methodology and are acceptable.
In Section 3/4.11.C, the licensee proposed minor editorial changes to the second paragraph. These changes affect the first two sentences and are as follows:
To assure that the fuel cladding integrity Safety Limit is not exceeded i
during any anticipated abnormal operational transient, the most limiting transients are analyzed to determine which result in the largest reduction in the CRITICAL POWER RATIO (CPR). The type of transients evaluated are change of flow, increase in pressure and power, positive jl reactivity insertion, and coolant temperature decrease.
Furthermore, the fourth paragraph is replaced with the following which again clarifies the SPC methodology which uses four scram insertion points to calculate MCPR Operating Limit and MCPR Safety Limit:
MCPR Operating Limits are presented in the CORE OPERATING LIMITS REPORT (COLR) for both Nominal Scram Speed (NSS) and Technical Specification Scram Speed (TSSS) insertica times. The negative reactivity insertion i
rate resulting from the scram plays a major role in providing the required protection against violating the Safety Limit MCPR during transient events. Faster scram insertion times provide greater protection and allow for improved MCPR performance. The application of NSS MCPR limits takes advantage of improved scram insertion rates, while the TSSS MCPR limits provide the necess:;ry protection for the slowest allowable average scram insertion times identified in Specification 3.3.E.
The measured scram insertion times are compared with the nominal scram insertion times and the Technical Specification Scram Speeds. The appropriate operating limit is applied, as specified in the COLR.
For core flows less than rated, the MCPR Operating Limit established in the specification is adjusted to provide protection of the Safety Limit MCPR in the event of an uncontrolled recirculation flow increase to the physical limit of the pump.
Protection is provided for manual and automatic flow control by applying the appropriate flow dependent MCPR
. limits presented in the COLR. The MCPR Operating Limit for a given power / flow state is the greatest value of MCPR as given by the rated conditions MCPR limit or the flow dependent MCPR limit.
For automatic flow control, in addition to protecting the Safety Limit MCPR during the flow run-up event, protection is provided to prevent exceeding the rated flow MCPR Operating Limit during an automatic flow increase to rated core flow.
1 The proposed change appropriately clarifies how Dresden will use the NRC-approved SPC methodology and does not change the current requirement that MCPR meet the limits specified in the COLR. Therofore, the proposed change to the
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TS bases is acceptable.
2.9 Reactor Core TS 5.3.A Fuel Assemblies, provides a description of the fuel assemblies.
In Amendment Nos. 159 and 154 for Units 2 and 3, respectively, the NRC added footnotes to only allow operation in Modes 3, 4, and 5 with ATRIUM-98 fuel other than lead test assemblies. With Comed's consent, the footnotes in this section are being modified. The modification to the footnotes will allow Dresdea, Unit 3, Cycle 15 to operate in all modes with ATRIUM-9B fuel. The staff is reviewing and approving the MCPR calculation for ATRIUM-9B fuel on a cycle-specific basis, as described below, until Appendix D of ANF-1125(P),
Supplement 1, is approved by the staff and adM to the Dresden TS.
- Dresden, Unit 2, operation is still limited to Modes 3, 4 and 5 with ATRIUM-9B fuel other than lead test assemblies.
l MCPR Safety Limit Calculation Based on the findings in the March 1997 audit at SPC, the value of 0.01 for the additive constant uncertainty (ACU) originally d?termined for the ATRIUM-98 fuel in Dresden, Unit 3, has been found to be based on an inadequate data base. A new estimate of the ACU for ATRIUM-9B fuel has been developed and submitted to NRC for review in a new topical report, Appendix D of ANF-1125(P), Supplement 1.
The topical report is currently in the review process, but Comed is seeking approval of an estimated value of the ACU for Dresden, Unit 3, that can be used in the interim. The proposed interim estimate of the ACU is obtained as follows:
s a.
first, determine the difference between the original ACU (0.01) and the estimated ACU (0.0195) obtained by the new methodology proposed in Appendix D of ANF-1125(P), Supplement 1.
This results in a delta-ACU of 0.0095; and b.
on the assumption that this delta-ACU represents a reasonable measure of the error in the original ACU (0.01), SPC proposes to obtain a conservative estimate of the error by multiplying the delta-ACU by two.
A new estimate of the ACU is obtained by adding this conservative estimate of the error to the original ACU for the ATRIUM-9B fuel in
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4 Dresden, Unit 3; that is, the new conservative estimate of the ACU is 0.029.
The resulting conservative ACU of 0.029 for ATRIUM-9B fuel was used to calculate the Safety Limit for MCPR (SLMCPR) for Dresden, Unit 3, Cycle 15.
SLMCPR increased from 1.05 and 1.06 for two-loop and single-loop operation, j '
respectively, in the May 1996 calculation to 1.08 and 1.09 for two-loop and single-loop operation, respectively, with an ACU of 0.029. Comed provided i
additional information in a letter dated June 10, 1997, with respect to the l
staff concern about the conservatism of the result (e.g. 0.0997% of fuel rods in boiling transition for two-loop analysis). The staff has reviewed this 3
information and found that the justification for a MCPR of 1.08 for two-loop I'
operation is acceptable.
Because 1.08 and 1.09 are consistent with the current Dresden TS 2.1.B no change to the TS is necessary for Unit 3, Cycle 15.
The ACU estimate of 0.029 for Dresden, Unit 3, is an increase of 190 percent i
over the original ACU of 0.01, and an increase of approximately 49 percent over the estimated ACU for ATRIUM-9B fuel that was determined in Appendix D of i
ANF-1125(P), Supplement 1.
The ANFB correlation fits the available ATRIUM-9B j
data sets with adequate margin to support the assumption that this interim estimate of the ACU (0.029) is conservative.
Further, it is also reasonable i
i to assume that this interim estimate will be found to b.e conservative even if the ACU value of 0.0195 determined in Appendix D of ANF-1125(P), Supplement 1, must be modified as a result of findings of the review of the submittal.
j Therefore, the proposed safety limit for minimum critical power (SLMCPR) of 1.08 is acceptable based on the interim proposed ACU of 0.029. However, SPC and Comed have committed to revise their SLMCPR calculations once Appendix D of ANF-1125(P), Supplement 1, is reviewed and a final value for the estimated ACU is approved.
2.10 Reactor Coolant System Section 182a of the Atomic Energy Act of 1954, as amended, requires applications for nuclear power plant operating licenses to include TS as part of the license. The Commission's regulatory requirements related to the content of TS are set forth in 10 CFR 50.36. That regulation requires that the TS 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 TS in its " Final Policy Statement on Technical Specification Improvements for Nuclear Power i
Reactors" 58 FR 39132 (July 22, 1993), in which the Commission indicated that compliance with the Final Policy Statement satisfies Section 182a of the Act.
In particular, the Commission indicated that certain items could be relocated from the TS to licensee-controlled documents, consistent with the standard 4
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enunciated in Portland General Electric Co. (Trojan Nuclear Plant), ALAB-531, l
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 i
reactor operation is deemed necessary to obviate the possibility of an i
abncrm:1 situation or event giving rise to an immediate threat to the public j
health and safety."
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The Final Policy Statement identified four criteria to be used in determining i
whether a particular matter. is required to be included in the TS limiting l
conditions for operation, is follows:
(1) installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary; (2) a process variable,
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j design feature, or operating restriction that is an initial condition of a l
design-basis accident or transient analysis that either assumes the failure of j
or presents a challenge to the integrity of a fission product barrier; (3) a structure, system, or component that is part of a primary success path and
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which functions'or actuates'to mitigate a design-basis accident or transient l
that either assumes the failure of or presents a challenge to the integrity of i
a fission product barrier; (4) a structure, system, or component which operating experience or probabilistic safety assessment has shown to be i
significant to public health and safety. As a result, existing TS l
requirements which fall within or satisfy any of the criteria in the Final j
Policy Statement must be retained in the TS, while those TS requirements which do not fall within or satisfy these criteria may be relocated to other i
licensee-controlled documents. The Commission amended 10 CFR 50.36 in j
July 1995, to codify and incorporate these four criteria.
i j.
TS 5.4 describes the design pressure, temperature, and volume of the reactor i
coolant system. The licensee proposed to relocate the contents of Specification 5.4 to the UFSAR.
Page 5-6 and Table of Contents page XIV are j
modified to read, "[ INTENTIONALLY BLANK)."
l Design temperatures, pressures, and volumes of the Reactor Coolant System in i
i existing TS Section 5.4 will be detailed in the Updated Final Safety Analysis Report (UFSAR).
Changes to these facility design parameters are controlled by the requirements of 10 CFR 50.59.
Furthermore, these design parameters are encompassed by existing TS Limiting Condition for Operation (LCOs) that establish acceptable requirements for ensuring that performance of the reactor coolant system is maintained. Any changes to the LCOs would receive prior NRC review and approval. Since the features with a potential to impact safety are sufficiently addressed by LCOs, and since design features, if altered in accordance with 10 CFR 50.59, would not result in a significant impact on safety,-the criteria of 10 CFR 50.36(c)(4) for including the above design features in the TS are not met.
The above relocated requirements relating to design features are not required to be in the TS under 10 CFR 50.36, and are not required to obviate the possibility of an abnormal situation or event giving rise to an immediate i
h,
threat to public health and safety.
In addition, the staff finds that sufficient regulatory controls exist under 10 CFR 50.59 to assure continued protection of public health and safety. This proposed change is consistent with Improved Standard Technical Specifications, NUREG-1433, Revision 1, and is acceptable. The Additional Condition in Appendix B of the license states the acceptability of the removal of the contents of Specification 5.4.
Accordingly, the staff has concluded that these requirements may be relocated from the TS to the licensee's UFSAR.
2.11 Reportina Reauirements TS 6.9 requires that in addition to the applicable reporting requirements of Title 10, Code of federal Regulations, the identified reports shall be submitted to the Regional Administrator of the appropriate Regional Office of the NRC unless otherwise noted. TS 6.9.A.6.a(4) describes the MCPR limit in i
the COLR. The licensee proposed to delete the 20 percent in the statement i
" including 20 percent scram insertion time" to reflect the SPC methodology.
The proposed change will state " including scram insertion times." This reflects the current SPC methodology and is acceptable.
TS 6.9.A.6.b lists the analytical methods used to determine the operating limits that are previously reviewed and approved by the NRC in the latest approved revision or supplement of topical reports. The licensee proposed to modify the list with additional topical reports. The additional topical reports are those used in SPC methodology and have been approved by the NRC for use at Dresden. The staff finds this change acceptable because the use of identified NRC-approved methodologies will ensure that the values for cycle-specific parameters are deterr.ined consistent with applicable design bases and safety limits, and assist safe operation of the facility.
1 3.0 Summary Comed requested changes to the Dresden Nuclear Power Station, Units 2 and 3, TS which would incorporate NRC-approved thermal limit licensing methodology in the list of approved methodologies used in establishing the cycle-specific
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thermal limits. Other minor editorial changes were also proposed. The staff concluded that these TS revisions are compatible with the STS.
The SPC methodology, with the modified ACU as described above, for ATRIUM-9B fuel-other than lead test assemblies is approved for Dresden, Unit 3, Cycle 15 only. This methodology was not approved for Unit 2 because the licensee has not addressed the use of a higher ACU for Dresden, Unit 2.
Based on the above, the staff concluded that operation in the proposed manner will not endanger the health and safety of the public and the issuance of the amendments will not be inimical to the common defense and security or to the l
health and safety of the public.
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4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendments.
The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to the 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, 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 consider & tion, and there has been no public comment on such finding (62 FR 17227). 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 environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 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 manner, (2) such i
activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: K. Kavanagh S. Bailey T. Huang Dated: June 12, 1997