1CAN038914, Application for Temporary Amend to License DPR-51 to Limit Max Operating Power Level to 1,915 Mwt.Fee Paid
| ML20247D409 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/23/1989 |
| From: | Tison Campbell ARKANSAS POWER & LIGHT CO. |
| To: | Calvo J NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| 1CAN038914, 1CAN38914, NUDOCS 8903310073 | |
| Download: ML20247D409 (19) | |
Text
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l ARKANSAS POWER & LIGHT COMPANY CAPIT0L TOWER BUILDING /P. O. BOX 551/LITTLE ROCK ARKANSAS 72203/(501) 377 3525 T. GENE CAMPBELL ad M, M Wee President - Nuclear j l
1CAN038914 Document Control Desk Mail Station P1-137 Washington, DC 20555 ATTN: Mr. Jose A. Calvo, Director Project Directorate, Region IV Division of Reactor Projects III, IV, V and Special Projects
SUBJECT:
Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 Request for Emergency License Amendment (Temporary)
Dear Mr. Calvo:
As discussed in a conference call with your staff and the NRC Region IV staff on March 20, 1989, AP&L recently identified a new postulated small break for ANO-1 not bounded by our current small break loss of coolant (LOCA) analyses. This condition was reported to the NRC in a 10CFR50.72 report on March 18, 1989. AP&L has been aggressively working with the ANO-1 nuclear steam supply system vendor, Babcock & Wilcox, since discovery of this postulated condition to perform analyses nccessary to establish appropriate measures to assure operation within the bounds of applicable ECCS requirements, upon restart of ANO-1. As discussed briefly below and described in detail in the attached license amendment request, these analyses demonstrate that power operation, in accordance with applicable emergency core cooling system (ECCS) criteria in 10CFR50.46 and Appendix K, can be maintained with adequate margins of safety up to and at a power level of 1915 megawatts thermal (74% of full power). Accordingly, AP&L submits with this letter (see attachment), a request to amend temporarily the ANO-1 license to reflect this restriction. This amendment is requested only until a permanent modification permitting resumption of full power operation is approved and implemented. Appropriate temporary changes to plant operating procedures and administrative controls for maximum power level and the Reactor Protection System high power trip setpoint will also be made.
A$I 8903310073 890323 PDR ADOCK 05000313 1
(9 P PDC p//cggk 0}-Y MEMBEA MOOLE SOUTH UTILITIES SYSTEM
. ICAN038914 March 23, 1989 ANO-1 is currently in a maintenance outage which is scheduled to be completed on March 26, 1989. Our current schedule would allow for heat-up i to begin the morning of March 26th, with criticality scheduled for the evening of March 27, 1989. In order to prevent a delay in plant heatup beyond 350 F (Technical Specification 3.3.2, High Pressure Injection System operability requirements), AP&L *equests that the attached proposed change be processed in accordance with the " emergency" provisions of 10CFR50.91(a)(5).
Application of 10CFR50.91(a)(5) is warranted because AP&L was unaware of the consequences of this new limiting break before March 18, 1989, and thus could not avoid this situation. Prior to that time, the previous LOCA '
analyses were considered to be bounding. Further, AP&L promptly undertook measures to define an appropriate response to this newly identified condition, and has prepared and submitted this request as soon thereafter as practical. Also, as noted above, this emergency change is needed in order to permit timely resumption of operation and to prevent an extended outage pending completion of the normal amendment process. Finally, in accordance with 10CFR50.91(a)(1), and using the criteria in 10CFR50.92(c), AP&L has determined that this change involves no significant hazards consideration.
The request for amendment and detailed bases are set forth in the attached submittal. As noted therein, should the NRC Staff require, AP&L requests that a temporary waiver of compliance be granted to permit plant startup
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pending completion of the staff formal review and approval of the requested amendment. Further upon receipt of the temporary waiver, AP&L will conservatively limit power to 50% maximum rated power pending completion of the formal review of the amendment request.
Also, in accordance with 10CFR50.91(b)(1), a copy of this amendment request and attachment has been sent to Ms. Greta Dicus,. Director, Division of Radiation Control and Emergency Management, Arkansas Department of Health.
A check in the amount of $150.00 is included herein as an application fee in accordance with 10CFR170.12(c).
Very truly yours, -
Y D T. Gene Campbel TGC:as Attachments cc w/ attachments: Ms. Greta Dicus, Director Division of Radiation Control and Emergency Management Arkansas Department of Health 4815 West Markham Street Little Rock, AR 72201 i
___---------,----,_.--_-a.----- - , _ - - - _ _ - - _ _ - - , - - _ - _ _ - - , - - - - _ - , - - - - - - . - - - - . . . - - _ _ - , _ . - - - - - - - - _ - . - . _ . . - -
STATE OF ARVANSAS )
) SS COUilTY OF PULASKI )
I, T. Gene Campbell, being duly sworn, subscribe to and say that I am Vice President, Nuclear for Arkansas Power & Light Company; that I have full authority to execute this oath; that I have read the document numbered ICAN038914 and know the contents thereof; and that to the best of my knowledge, information and belief the statements in it are true.
/
WJ T. Gene Campb SUBSCRIBED AND SWORN T0 before me, a - Notary Public-in and- for- the-- ---- - - -
County and State above named, this 8 day of M b ,
1989.
I as/L Notary Public l
l My Commission Expires:
V-/Q-$ 9 l
l
ATTACHMENT 1 PROPOSED LICENSE CHANGE
,. REQUEST FOR EMERGENCY AMENDMENT l (TEMPORARY)
ARKANSAS POWER & LIGHT COMPANY ARKANSAS NUCLEAR ONE, UNIT 1 DOCKET NO. 50-313 MARCH 23, 1989 INTRODUCTION Arkansas Power & Light Company (AP&L) herein requests a temporary amendment to the Arkansas Nuclear One, Unit 1 (AND-1) Operating License. This amendment will limit the maximum operating power level to 1915 megawatts thermal (74% of full power operation). This change is necessary to assure that adequate core cooling will be available in the event of a newly postulated small pipe break in the High Pressure Injection (HPI) line, just upstream of the Reactor Coolant System (RCS) cold leg connection. Because this change is necessary to permit scheduled plant startup, it is requested on an emergency basis. AP&L only recently discovered the potential implication of this break and, thus, could not have reasonably prevented the emergency situation. The requested change and associated bases are described fully below. AP&L notes that the amendment will be temporary, i.e., effective only until a permanent modification is approved and implemented to permit resumption of full power operation.
BACKGROUND Identification of Condition On January 20,1989, ANO-1 experienced a reactor trip initiated by a generator lockout. Following the trip, certain feedwater control system and electrical distribution system problems required the operators to manually initiate additional HPI flow to the RCS. It was later discovered that the check valve in the "B" HPI injection line had failed to reseat after HPI flow was terminated. This allowed reactor coolant to flow into the HPI line resulting in the line being overheated.1 As a result of the January 1989 transient, AP&L undertook a thorough review of the HPI system. This review included a reevaluation of the qualification and ability of both the individual components and the HPI system as a whole to withstand all conditions that could result from transients and steady state operations. During this review, it was discovered that a postulated break of an HPI injection line, just upstream of the RCS cold leg connection and downstream of the first check valve, could constitute a small break LOCA not currently enveloped by the approved 10CFR50.46 and Appendix K analyses.
AP&L requested that Babcock & Wilcox (B&W), the nuclear steam supply system vendor for ANO-1, evaluate the impact of this postulated break or, current ECCS evaluations. B&W analyzed the break and inforned AP&L that the ISee letter from Mr. T. G. Campbell to Mr. Jose A. Calvo, February 19, 1989 (ICAN028909) 1
l postulated break did not appear to be enveloped by previously postulated i
br,eaks and that the ANO-1 HPI system might not be able to provide adequate l
. core cooling (using conservative Appendix K assumptions) should the break occur at high power operation. AP&L promptly reported this finding pursuant .
to 10CFR50.72 on March 18, 1989. B&W thereafter undertook to define the operating parameters for which this postulated break may not be addressed with the' current ECCS response capabilities at ANO-1. B&W determined that for power operation up to 74% of full power, the current ECCS response using the HPI system would provide adequate core cooling in the event the postulated break were to occur.
Current ECCS Design Basis The ANO-1 ECCS is designed to meet the criteria of 10CFR50.46 and 10CFR, Part 50, Appendix K. AP&L uses the B&W proprietary CRAF1 1 code to analyze ECCS performance at ANO-1 (B&W topical reports BAW-10103 and 10104). AP&L notes that the use of the CRAFT code and the ECCS design assumptions set forth in 10CFR, Part 50, Appendix K provides significant conservatism in the ECCS response analyses. These conservatism have also been clearly recognized by the NRC and have led to recent revisions to Appendix K that allow methods to reduce the significant levels of conservatism (reference NUREG-1230). Specifically, the NRC acknowledged that restrictions on operation due to ECCS criteria (i.e., 10CFR50.45) are more stringent than necessary. In particular, the conservatism relate to assumed decay heat behavior, critical flow, metal-water reaction, and many other phenomenological considerations. Nevertheless, AP&L has employed the existing conservative analytical criteria in evaluating the present condition, rather than the NRC-endorsed realistic "best estimate" analyses for evaluating ECCS performance. AP&L believes that with the use of appropriate best estimate calculations, the current postulated condition would be shown to be addressed by the current ECCS HPI design. (For example, AP&L has performed preliminary analyses which show that merely using a more realistic decay heat curve (1.0 x ANS) the core remains covered for this break.) Nonetheless, AP&L seeks consideration of the proposed response to this condition using the current, highly conservative ECCS licensing basis for ANO-1.
Single Failure Standard Also pertinent to this amendment request are the standards establishing single failure requirements for ECCS evaluations. Provisions addressing consideration of single failures are set forth in 10CFR Part 50, Appendix A, General Design Criterion 35 (GDC 35). GDC 35 provides an applicable part as follows:
Suitable redundancy in components and features...shall be provided to assure that for on-site electric power system operation (assuming off-site power is not available) and for off-site electric power system operation (assuming on-site power is not available) the safety function can be accomplished, assuming a single failure.
AP&L has considered appropriate single failures in its analysis of this condition. Accordingly, AP&L has pursued and proposes herein a change to ANO-1 license which assures continued satisfaction of current ECCS standards, including associated single failure criteria.
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DISCUSSION
. High Pressure Injection System The HPI system is a part of both the makeup and purification system and the emergency core cooling system. It uses two of the three makeup pumps for injection of coolant from the Borated Water Storage Tank for core cooling.
The HPI system consists of two redundant trains. A general layout of the HPI system is provided in the attached Figure 1. The HPI system is designed to prevent uncovering of the core for small break LOCAs, where high RCS pressure is maintairad, and delays uncovering of the core for intermediate size breaks. With respect to small breaks each train is designed to provide sufficient coolant to compensate for breaks up to four inches in diameter. For more detailed information on the HPI system, see Chapter 6 of the ANO-1 SAR.
Postulated Small Break Scenario The particular postulated break involved in this request is a small break LOCA, less than .03 ft.2, located in any one of the HPI injection lines, just upstream of the RCS cold leg connection and downstream of the first check valve (see Figure 1). For this postulated break and assuming a worst case single failure (i.e., of one of the diesel generators, rendering one of the HPI pumps inoperable), virtually all HPI flow would exit the break with none reaching the core. This would occur because backpressure on the broken line would essentially be containment pressure while the other three lines (via cross-connects) would see RCS pressure of >1000 psig. This situation would continue until the operator took action to balance the indicated flows.
The motor operated valves used to throttle flow in the HPI lines have been demonstrated capeble of closing against maximum HPI pump output pressure.
With such action, 50% of the flow would reach the core (see Figure 2). This scenario results in essentially 0% HP' ' low until operator action to balance flow (modeled at ten minutes following the break) and then 50% HPI flow thereafter. However, that response is not bounded by the existing ECCS analysis, which assumes 50% flow from one HPI pump until operator action to balance flow (also at 10 minutes) and then 70% flow to the core from one HPI pump for the remainder of the accident.2 2 The requirement for balancing flows, rather than isolating the high flow or apparently affected line, addresses the concern of a HPI line i break and subsequent line pinch. If the highest flow injection line was simply isolated following a HPI line break and postulated line pinch, it is possible that line may be the intact line and the HPI flow that was reaching the RCS would thus be terminated.
AP&L also notes that installation of the HPI cross-connects was undertaken in 1978 to address a break that was postulated on the bottom of one of the cold legs at the RCP discharge, in which case all the HPI flow from one line went into the cold leg and out the break, bypassing the core. Without the cross-connects and assuming a single failure rendering the unaffected HPI train inoperable, only 50% of the flow from one HPI pump would reach the core. This reduced flow was insufficient to protect the core at 100% power, so the cross-connects were installed to permit balancing of the flows.
3
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. AP&L reiterates that due to the_ conservatism of the 10CFR50 Appendix K calculati,ons, the extremely low probability'of this specific break
. accompanied by a loss of off-site power, and an accompanying diesel failure, this previously unanalyzed condition does not represent a substantial safety hazard for. realistic assumptions. Nonetheless, as discussed below, AP&L proposes an approach below that fully adopts and retains the applicable licensing basis conservatism.
Proposed Interim Response To address this issue, AP&L explored several options. These options included permanent hardware fixes, revisions to ECCS analyses using "best-estimate" assumptions, and sufficient limitation of power level to assure compliance with 10CFR50.46 using current licensing basis analyses and assumptions. Although permanent hardware fixes (as discussed below) will be examined as long-term responses to permit resumption of full power operation, AP&L concluded that significant additional outage time would be required to develop designs, procure equipment and make necessary hardware modifications at this time. Further, revising analyses to use the.
"best-estimate" methodology endorsed by the NRC 'or pursuing any necessary exemptions from 10CFR50.46 would require extensive reanalysis and NRC review. This option, like the first, could result in significant additional outage time. Accordingly, AP&L decided to pursue limiting power sufficiently to' assure continued compliance with 10CFR50.46 as an interim response. This approach, as well as being highly conservative, would !
support the current restart schedule, although restricting operation to a reduced power level.
In accordance with this decision, AP&L requested B&W to evaluate the new small break scenario, and to ascertain any operating restrictions necessary to maintain ANO-1 compliance with ECCS requirements. B&W undertook a comparative analysis of the ANO-1 system with a representative analysis for another B&W reactor. Specifically, an HPI line break analysis had been performed by B&W for the Consumers Power Company Midland Unit 1 and 2 plants. This analysis has been reviewed and approved by the NRC staff.
(These plants were typical of the 177 FA Lowered Loop design.) The initial core power level for this analysis was selected to be representative of the B&W plant with the highest power level. 2772 MWt (as compared to ANO-l's 2568 MWt), and included 10CFR50.46 and Appendix K methods and assumptions.
Further, this analysis used an ECCS system specific to the Midland design ,
such that a significant portion of the HPI flow was injected into the core l without operator action. In addition, 20 minutes following initiation of HPI, the broken HPI line was isolated. After this time, 100% of the HPI flow was injected into the RCS. The Midland analysis showed that at approximately one hour into the transient, adequate HPI flow was achieved to match the core boil-off rate and initiate RCS refill, thus establishing the path to long-term cooling.
B&W then performed a comparative evaluation of the Midland HPI line break transient assuming the AN0-1 HPI configuration. ANO-1 operator guidance is being revised to require the operator to balance the flow in all injection lines rather than isolate the break as assumed in the Midland analysis. B&W credited successful flow balancing at 10 minutes, resulting in essentially zero HPI injection prior to 10 minutes and 50% HPI injection after 10 minutes (50% of the HPI flow is lost out of the line break).
4 E _. _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _
Further, operation at 100% power (2772 MMt (Midland) and 2568 MWt (ANO-1))
was comparatively evaluated to account for potential differences between the !
- Midland and the ANO-1 response. This evaluation indicated that were this l small break to occur during operation at full power core uncovery could result. Therefore, full power operation of ANO-1 could not be justified.
In addition, conservative evaluations assuming 1.0 times the 1971 ANS 5.1 decay heat standard (compared to the 1.2 multiplier required by 10CFR Part 50, Appendix K) and an initial power level of 2568 MWt were shown to be ,
acceptable (i.e. core uncovery was not predicted). RCS refill was estimated ]
to occur at approximately 3300 seconds.
In addition, an evaluation to assess operation of ANO-1 at reduced power, I applying the full 10CFR50.46 and Appendix K requirements (including the 1.2 ANS decay heat factor), was made. This evaluation demonstrated that operation at or below 74% full power was acceptable in that no core uncovery would occur, and RCS refill is achieved within one hour into the transient.
The detailed calculation performed by B&W and reviewed and approved by AP&L !
in support of this analysis will be transmitted directly by B&W under ;
separate cover. (B&W 1etter from J. H. Taylor to J. A. Calvo,
Subject:
ANO-1 HPI Line Break Evaluation.)
B&W and AP&L also evaluated the impact of reduced power level upon the various Safety Analysis Report (SAR) accident analyses and upon the various fuel related analyses performed as part of the reload methodology. The accident analyses are all bounded by either the full power or zero power cases documented in SAR Chapter 14. The additional considerations of containment analyses and High Energy Line Break (HELB) analysis were also addressed and shown to be acceptable. No impact was identified on the various fuel-related analyses for operation of an additional 50 effective full power days (EFPD) under restricted conditions. AP&L will perform specific reanalysis to determine any impact upon Technical Specification limits for operation beyond an additional 50 EFPD, and propose Technical Specification changes if appropriate under the provisions of 10CFR50.91(a)(2)-(4).
AP&L has also evaluated the existing HPI flow instrumentation to assure it is appropriately classified per RG 1.97, based upon the need to use it for balancing injection flow in the various HPI line break / pinch scenarios. The existing configuration (Type D, Category 2 variable) is considered adequate based on the low probability of the event of concern and the inherent margin in the analyses as described above. The existing Category 2 instrumentation differs from Category 1 only in redundancy and continuous real-time recording (recording capability does exist on the SPDS).
Potential Permanent Response AP&L is examining possible permanent modifications that would permit resumption of full power by operation addressing the postulated HPI break consistent with the present ECCS licensing basis. AP&L intends to finalize a proposed permanent modification as soon as practical. Implementation of that modification will occur at the first outage of sufficient duration to permit the change following NRC staff approval. At that time, the condition for resumption of full power operation contained in the temporary amendment presently requested will have been satisfied, and procedural or operating changes will be implemented to reflect resumption of full power operation.
5
~, ".The permanant responses AP&L is presently considering include:
. 1. Cavitating venturis - these devices are designed to limit maximum flows through each lines while not limiting flows under less restrictive flow conditions.
- 2. HPI pump discharge cross-connects - this approach would involve adding motor operated valves on the HPI discharge header isolation valves. The design would provide for the HPI flow split between injection legs such that a 70/30 flow balance could be achieved following operator action. This improvement in HPI flow is estimated adequate to resolve the HPI line break concern.
- 3. Enhanced indication - several possibilities involving additional instrumentation are being considered. The instrumentation could potentially be used to support additional operator action (e.g.,
sufficient to allow identification and isolation of the broken HPI line).
AP&L will keep the staff apprised of its decision regarding permanent responses.
Review of Potential Similar Situations AP&L also undertook a reexamination of ECCS system design to assess the possibility of other unanalyzed break scenarios in similarly configured ECCS flow paths. The only other similar injection path identified in that review is the common LPI/ Core Flood Tank injection line. Adequate protection against breaks in that line is provided by flow limiting orifices in the injection nozzles (these limit flow from the reactor vessel) and cavitating venturis in the LPI injection lines. These preclude similar LPI run-out conditions given a break in one of the injection lines. A break in the Core Flood line is protected by check valves isolating RCS blowdown and by the other fully redundant and independent Core Flood Tank. AP&L presently believes that other connections to the RCS are adequately addressed by isolation valve design provisions and potential breaks are enveloped by the current SBLOCA analyses.
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1 . .
ENERGENCY OPERATING LICENSE AMENDMENT I
Pursuant to 10CFR50.91(a)(5), AP&L hereby requests NRC approval on an emergency basis to amend Operating License DPR-51 for the ANO-1 Plant.
Emergency authorization is required so as not to prevent resumption of ;
operation of the facility as a result of restrictions imposed by the {
Technical Specifications, pending approval of this request. The plant is l presently shutdown and scheduled to reach criticality the evening of March i 27, 1989.
AP&L recognizes that this situation involves a condition for which present accident response capabilities are potentially insufficient to satisfy 10CFR50.46 and Appendix K requirements over the full range of authorized power operation, using current accident modelling techniques for Unit 1. ,
Accordingly, either an exemption from the present ECCS performance '
requirements, or an amendment as described herein appear necessary. The requested amendment is the preferred approach because it will maintain operation within the approved design criteria and applicable regulations.
l In addition, the requested emergency authorization is warranted because it i involves no significant hazards consideration, as demonstrated below.
Further, the present situation could not have been avoided. As noted above, AP&L promptly notified the NRC upon determining the potential significance of this condition and has pursued an expeditious resolution of the matter, including comprehensive measures to identify and confirm appropriate short and long-term responses. In the event this approval may not be received by that time, AP&L requests a temporary waiver of the particular Technical Specification provisions governing operability of the HPI system to permit operation of up to 50% power3pending receipt of the requested amendment.4 3 AP&L has voluntarily decided that, upon receipt of any necessary temporary waiver of compliance, it will conservatively hold power operation below 50% power pending approval of the requested amendment.
This level provides a substantial additional margin of safety for plant operation pending the staff's review of the proposed 74% power level restriction set forth in the amendment request.
4 Consistent with EGM-85-05B (Memorandum for the Regional Administrators from H. R. Denton, J. M. Taylor, " Relief from Technical Specification LC0's", February 17, 1987), this temporary waiver will be required only until approval of the requested operating license approval.
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1 D(SCRIPTION OF PROPOSED LICENSE CHANGE The specific amendment AP&L proposes involves Section 2.1.1, Maximum Power Level of of the ANO-1 Facility Operating License. In this section of the license, the maximum steady state reactor core power level would be temporarily reduced from 2568 megawatts thermal to 1915 megawatts thermal.
The change also provides that resumption of full power operation will be permitted upon approval and implementation of a proposed permanent modification. A copy of the proposed amendment is attached. Further, although not involving license changes, AP&L is implementing administrative controls to ensure that the restriction to 74% is carried out. I Specifically, to preclude operation above this power level, ANO-1 operating f i
procedures will be modified prior to restart to administratively limit power operation to 1915 megawatts thermal. Finally, although current safety l analyses which credit the high power trip remain enveloped at reduced power I levels, AP&L will reduce the trip setpoint to a setting of 83% to provide additional available margin for those trip scenarios.
i 8
(
BSSES FOR PROPOSED NO SIGNIFICANT. HAZARDS CONSIDERATION DETERMINATION In accordance with 10CFR50.92, AP&L has assessed whether the proposed change j involves a significant hazards consideration. AP&L has concluded that the j proposed change to limit operation to 74% of full power does not involve a significant hazards consideration because operation of Arkansas Nuclear One, '
Unit-1 in accordance with this change would not:
(1) Involve a significant increase in the probability or consequences l of an accident previously analyzed. j l
First, this change does not alter the probability of any i previously analyzed accident occurring. The change merely addresses a particular accident scenario without impacting accident-initiating events. Further, this proposed change will l not adversely affect the consequences of accidents which have been I previously analyzed. Any effect on previously analyzed accidents will be positive as the reactor will trip from a lower maximum power level as a result of the proposed change. ]
Further, the proposed change does not adversely affect the probability or consequences of the postulated HPI small break at issue here. Upon implementation of the proposed change, this I break will be fully addressed by available ECCS mechanisms f consistent with applicable ECCS requirements. In fact, this l response capability might not have been available absent the change under the present conservative licensing basis assumptions.
Overall, therefore, this change will neither reduce nor adversely impact the probability or consequences of accidents previously ,
analyzed. 1 i
(2) Create the possibility of a new or different kind of accident from '
any previously analyzed.
First, the ECCS response to other previously postulated accidents remains unchanged and within previously assessed limits of flow paths and flow distributions. Further, all systems and ECCS coolant delivery mechanisms remain, respectively, within their applicable performance limits and flow delivery capabilities. l Thus, system and component performance is not adversely affected by this change, thereby assuring that design capabilities of those systems are not challenged in a manner not previously assessed so as to create the possibility of a new or different kind of accident. Further, the proposed change would result in a reduction of the maximum allowable power level. This would not create the possibility of a new or different kind of accident from any previously analyzed as the possibility of reactor trips at lower power levels had already been considered in previous accident analyses.
(3) Involve a significant reduction in the margin of safety.
The proposed change to the maximum operating power level would reduce the maximum power level at which any transient could occur. .
The response to transients terminated at lower power levels is j
9
~
typically mildar and more easily controlled than the response to transients terminated at a higher power level. Therefore, the l - reduction of the maximum permissible power level would not involve i a significant reduction in the margin of safety previously l provided and, in fact, would generally increase the margin. With respect to the HPI break at issue, the margin of safety provided I by reducing the maximum power level now reflects the margins provided by application of the conservative assumptions and analytical approaches of 10CFR50.45 and 10CFR Part 50, Appendix K.
Thus, the inherent margins of safety provided by those criteria now also are provided for this postulated break.
The Commission has prc ided guidance concerning the application of these j standards by providing examples of ctanges involving no significant hazards considerations . The proposed amendment most closely matches example (ii)
"A change that constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications."
Therefore, based on the above, AP&L has concluded that the proposed change does not involve a significant hazards consideration.
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LICENSE AMENDMENT REQUEST IN THE MATTER OF AMENDING LICENSE NO. DPR-51 ARKANSAS POWER & LIGHT COMPANY ARKANSAS NUCLEAR ONE, UNIT 1 DOCKET NO. 50-313 MARCH 23, 1989 1
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- c. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, So tion 30.34 of Part 30, Section 40.41
-of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 l
of Part 70;_is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor' core power levels not in excess of 1915 megawatts thermal, pending NRC approval and licensee implementation of a permanent modification to address a High Pressure Injection System small breck LOCA addressed in the licensee's letter dated March 23, 1989, at which time full power operation may resume at steady state reactor core power-levels not in excess of 2568 megawatts thermal.
(2) Technical Specifications Amendment l The Technical Specifications contained in Appendices A and
- 117 l B, as revised through Amendment No. 117 are hereby 03/10/89 l incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3) The' licensee may proceed with and is required to complete the modifications ~ identified in Paragraphs 3.1 through 3.19 of the NRC's Fire Protection Safety Evaluation (SE) on the facility dated August 22, 1978 and supplements thereto.
These modifications shall be completed as specified in Table 3.1 of the Safety Evaluation Report or supplements thereto.
In addition, the licensee may proceed with and is required to complete the modifications identified in Supplement 1 to the Fire Protection Safety Evaluation Report, and any future supplements. These modifications shall be completed by the dates identified in the supplement.
(4) Physical Protection The licensee shall fully implement and maintain in effect all l
provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingent,y i plans including amendments made pursuant to provisions of the ;
Miscellaneous Amendments and Search Requirements revisions to I 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled.
(a) " Arkansas Nuclear One Physical Security Plan," with revisions submitted through February 24, 1988; 12 >
.- c ,
- (b) " Arkansas Nuclear One Guard Training and'Quali?ication
.- Plan," with revisions submitted through August 20, 1985;
- and, (c) " Arkansas Nuclear One Safeguards Contingency Plan," with revisions submitted throegh May 30, 1980. 1 k
Changes made in accordance with~10 CFR 73.55 shall be 1 implemented in accordance with the schedule set forth. j therein.
(5) Systems Integrity' The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This program shall include the following:
- 1. Provisions establishing preventive maintenance and periodic visual inspection requirements, and
- 2. Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.
(6) Iodine Monitoring The licensee shall implement a program which will. ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.
program shall include the.following:
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- 1. Training of personnel,
- 2. Procedures for monitoring, and
- 3. Provisions for maintenance of sampling and analysis equipment.
(7) Secondary Water Chemistry Monitoring A secondary water chemistry monitoring program shall be implemented to minimize steam generator tube degradation.
This program shall include:
- 1. Identification of a sampling schedule for the critical parameters and control points for these parameters;
- 2. Identification of the procedures used to measure the values of the critical parameters;
- 3. Identification of process sampling points;
- 4. Procedures for the recording and management of data;
- 5. Procedures defining corrective actions for off-control point chemistry conditions; and 13 L___________-___________-______________________
.. .~ . _ _ . .- -_ _____ - _-
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- 6. A procedure id:ntifying the authority responsible for l
, the interpretation of the data and the sequ:nce and !
. timing of administrative events required to initiate a i corrective action.
- d. This license is effective as of the date of issuance and shall '
l expire at midnight, December 6, 2008.
FOR THE ATOMIC ENERGY COMMISSION 1 A. Giambusso, Deputy Director i l
Date of Issuance: May 21 1974 l
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