ML20202E557
| ML20202E557 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 02/10/1998 |
| From: | Sorensen P NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20202E516 | List: |
| References | |
| NUDOCS 9802180178 | |
| Download: ML20202E557 (9) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET Nos. 50 282 50 306 REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR 42 & DPR 60 LICENSE AMENDMENT REQUEST DATED February 10,1998 Qne Time Reductlpy of Mode 6 Reactor Coolant System Boron Concentration Northern Statee Power Company, a Minnesota corporation, requests authorization for changes to the Prairie Island Operating License, Appendix A as shown in the attachments labeled Exhibits A, B, and C. Exhibit A describes the proposed changes, reasons for the changes, and the supporting safety evaluation and significant hazards determination. Exhibit B contains current Prairie Island Technical Specification page Table TS.1-1 marked up to show the proposed changes. Exhibit C contains the revised Techniccl Specification page incorporating the proposed changes.
This letter and its attachments contain no restricted or other defense information.
NORTHERN STATES POWER COMPANY By I k-Jo@), Sorensen' Plant Manager, Prairie Island Nuclear Generating Plant On this day of W (64tL/ ofore me a t otary public in and for said County, personally appearedfi6hl P. Sorensen, Plant Manager, Prairie Island Nuclear Generating Plak3nd being first duly sworn acknowledged that he is authorized to execute this document on tshalf of Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the statements made in it are true and that it is not Interposed for delay. M s
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4 Exhibit A
, , Prairie Island Nuclear Generating Plant )
License Amendment Request Dated February 10,1998 Evaluation of Proposed Changes to the Technical Specifications Appendix A of Operating License DPR-42 and DPR 60 Pursuant to 10 CFR Part 50, Sections 50.59 and 50.90, the holders of Operating ;
Licenses DPR 42 and DPR-60 hereby propose the following changes to Appendix A, Technical Speelfications:
ggckaround and Reasons for Channes l Prairie Island Unit 2 was shut down on January 24,1998 to repair a small leak on the i
reactor coolant system. The unit was taken to the cold shutdown condition for the repair. The source of that leakage appeared initially to be the intermediate canopy seal weld for the part length control rod drive mechanism (CRDM) at location G9 on the reactor vessel head (see attachment). However, the canopy seal was covered and couldn't be directly observed. Filling and venting the reactor coolant system was initiated while the repair of the intermediate canopy ceal weld was still in progress.
During that fill and vent operation it was noted that water was leaking from the G9 part length CRDM approximately 1.5 inches above the intermediate canopy seal that was 1
being repaired.
' Because of the location of the affected CRDM, accessibils to the CRDM is limited. A video camera was utilized to investigate the leakage.. Examination of the leak location using the video camera identified water dripping from an apparent flaw in the wall of the CRDM Motor Tube Base above the intermediate canopy seal. Filling and venting was ,
secured and the reactor coclant system was drained to a level below the leak elevation.
Because the part length CRDMs are not used and are abandoned in place, it was decided that the part length CRDM at location G9, would be permanently removed.
This would eliminate the need for repair of the flaw and would facilitate the metallurgical evaluation of the flaw. Removal of the G9 part length CRDM was completed on February 8,1998. Preliminary plans are to also remove the remaining three part length CRDMs. However, it is possible that the other part length CRDMs could be left in place if it can be shown that no significant degradation exists.
Removal of the CRDMs requires capping of the associated reactor vessel head penetrations. Two options are available for capping of the penetrations. The preferred method is the installation of a cap which would be screwed onto the threaded end of the penetration and then seal welded, if the installation of screw on caps is found to be not feasible it will be necessary to utilize welded caps. To install a welded cap, the -
threaded portion of the penetration is removed and a cap is installed by welding it to the penetration. The second option is less desirable because of the limited
Exhibit A Pcg3 2 of 8 accessibility of the part length CRDM penetrations and the nood to remove the reactor vessel head to prevent foreign material from falling into the reactor coolant system during the cutting and welding process.
Prior to de-tensioning of a reactor vessel head closure stud and thus entry into Mode 6, Technical Specification Table TS.1-1 requires that the boron concentration of the reactor coolant system be sufficient to ensure that the more restrictive of the following conditions is met:
- a. K.n s 0.95, or
- b. Boron concentration a 2000 ppm.
In addition, we believe it is the intent of this Technical Specification requirement that the boron concentration be uniform throughout the reactor coolant system.
As discussed above, when Unit 2 was shutdown, it was believed that the reactor coolant leakage was from an intermediate canopy seal weld. These types of repairs have been performed in the past and it is not necessary to remove the reactor vessel head for such repairs. Because the need to remove the reactor vessel head and enter
- Mode 6 was not anticipated, the reactor coolant system was only borated to the concentration necessary to support Cold Shutdown (Mode 5) operations. Thus, the reactor coolant system boron concentration was measured at 1950 ppm when the unit reached cold shutdown and the reactor coolant pumps were stopped. The 1950 ppm boron concentration is not adequate to meet the requirements for entry into Mode 6. It follows that if it becomes necessary to enter Mode 6 to support resolution of the part lorsgth CRDM leakage it would be necessary to increase the reactor coolant system boron concentration to meet the requirements of the Technical Specification.
j Increasing the boron concentration uniformly throughout the reactor coolant system requires circulation of the coolant throughout the reactor coolant system. However, when in Mode 5, the reactor coolant pumps are secured and decay heat is being i removed via the residual heat removal (RHR) system. The RHR system provides forced circulation through the reactor vessel and fuel, but does not provide forced
! circulation through other portions of the reactor coolant system such as the steam l generators and intermediate legs, increasing reactor coolant system boron concentration uniformly to the level required for entry into Mode 6 would require operation of the reactor coolant pumps. There is no other method to ensure uniform mixing of boron concentration changes throughout the reactor coolant system and thus j meet the intent of the Technical Specification requirements. However, because the G9 part length CRDM has been removed from the reactor vessel head, the reactor coolant system is not considered intact and a reactor coolant pump cannot be started.
Due to concerns with the effect of non-uniform boron concentrations between the actise and inactive portions of the reactor coolant system, it is also not considered prudent to l
l
Exhibit A Paga 3 of 8 borate the active portion of the reactor coolant system to a concentration greater than that in the inactive portion.
Because of the inability to start the reactor coolant pumps and the concerns with borating only the active portion of the reactor coolant system, this License Amendment Request proposes a limited duration change to the Prairie Island Technical Specifications that would allow a reduction in the boron concentration required for Mode 6. The current requirement of greater than or equal to 2000 ppm would be reduced to greater 1,ian or equal to 1900 ppm if it becomes necessary to remove the reactor vessel head and enter Mode 6 during the current Unit 2 forced outage. If the proposed change were implemented, the active volume of the reactor coolant system would be maintained at a boron concentration of 21900 ppm to ensure that L remains below 0.95 under all postulated conditions.
The proposed change would only be valid for Mode 6 operations during the Unit 2 forced outage which began on January 24,1998.
Proposed Channes The proposed changes to the Prairie Island Technical Specifications are described below, and the specific wording changes to Technical Specifications are shown in Exhibit B.
This License Amendment Request proposes the addition of the following footnote to Technical Specification Table TS.1-1:
The reactor coolant system boron concentration shall be 21900 ppm for MODE 6 operation durin0 the Unit 2 forced outage which began on January 24,1998.
Even though the reactor coolant system boron concentration was measured at 1950 ppm when the reactor coolant pumps were stopped, a lower limit of 1900 ppm is being proposed to allow for 1% uncertainty in the measurement of boron concentration.
Safety Evaluation and Juatification of Chanaes Technical Specification Table TS.1 1 requires that the boron concentration of the reactor coolant system be sufficient to ensure that the more restrictive of the following conditions is met:
- a. s0.95,or
- b. Boron concentration 2 2000 ppm.
Based on a conservative Unit 2 core burnup up, and including penalties for calculational uncertainties, the reactor coolant system boron concentrations required to ensure M of 5 0.95 is 1634 ppm with all rods in at 68*F.
Exhibit A Pcge 4 ei' B This License Amendment Request proposes a limited duration change to Technical Specification Table TS.1 1 that would allow a reduction in the boron concentration required for Mode 6. The current requirement of greater than or equal to 2000 ppm would be reduced to greater than or equal to 1900 ppm if it becomes necessary to remove the reactor vessel head and enter Mode 6 during the current Unit 2 forced outage.
Por guidance provided in Revision 1 of NUREG-1431,' Standard Technical I Specifications, Westinghouse Plants', the rninimum beton concentration required while in Mode 6 ensures that a core K.n of s 0.95 is maintained during fuel handling operations. NUREG-1431 also states that plant procedures ensure the specified boron concentration in order to maintain an overall core reactivity of K., s 0.95 during fuel handling, with control rods and fuel assemblies assumed to be in the most adverse configuration (least negative reactivity) allowed by plant procedures. Based on this guidance it is clear that the intent of the Mode 6 boron concentration requirement is to provide additional shutdown margin during core alterations and fuel handling. This is supported by the fact that during Mode 5, when no fuel handling or core alterations are possible, K., only needs to be s 0.99.
If entry into Mode 6 is required for the part length CRDM repairs effort, it will only be necessary to remove the reactor vessei nead, no core alterations or fuel handling would be necessary and none are planned. There are two postulated events which could result in the addition of positive reactivity to the seactor core during the part length CRDM repairs: (1) the unplanned withdrawal of a control rod from the core during removal of the reactor vessel head, and (2) an inadvertent dilution of the reactor coolant system.
The inadvertent withdrawal of a control rod during a head lift has never occurred at Prairie Island. However, the minimum boron concentration needed to ensure that K.n will remain less than 0.95 with the highest worth rod fully withdrawn during the current Unit 2 outage has been calculated. That minimum boron concentration at 68'F and with uncertaintles applied is 1690 ppm. Thus, should a control rod be inadvertently withdrawn from the core during the head lift, the proposed aron concentration limit would provide at least a 210 ppm margin above the boron required to maintain K., s 0.95 under those conditions. In addition, because the reactor coolant system boron concentration was actually measured to be 1950 ppm, there would be additional margin above the proposed 1900 ppm limit. To address the unlikely possibility that more than one control rod were inadvertently withdrawn during the reactor vessel head lift, it was calculated that a reactor coolant boron concentrailon greater than 1625 ppm will ensure that the reactor remains suberitical even with all rods withdrawn at 68 F and with calculational uncertainties applied.
The analysis of a dilution accident during the current Unit 2 outage was performed.
That analysis, which conservatively assumed reduced inventory conditions and maximum possible dilution water flow, showed that a boron concentration greater than
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Exhibit A P ga 5 of 8
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1825 ppm will ensure that there is at least 24 minutes between the initiation of a chemical and volume control system malfunction accident (i.e. dilution), and the loss of all shutdown margin. This is longer than the time calculated in the Prairie Island USAR for a dilution accident during refueling. Therefore, the proposed boron concent.ation limit will ensure that the plant would remain bounded by the USAR analysis.
The boron concentrations noted above were calculated by the Northern States Power Nuclebr Analysis Department, using the NRC approved methods described in NSP topleal report NSPNAD-8101 A, Revision 1, Qualification of Reactor Physics Methods.
The calculations were based on a conservative core exposure and include a 1%
calculational uncertainty. All calculations were verified in accordance with NSP administrative controls.
During operation in Mode 6 for the current Unit 2 forced outage, the active volume boron concentration will be maintained at the required concentration (21900 ppm) which will include the RHR System and the upper regions of the vessel. Por Technical Specifications, sampling and chemical analysis of the reactor coolant system is required to be performed daily when in Mode 6. This sampling will be performed on the active volume of the reactor coolant system. If a decrease in the boron concentration occurs, actions would be taken to ensure that the concentration remains above the minimum required by the proposed Technical Specification requirements.
Conclusion As discussed above, the required Mode 6 boron concentration is !ntended to ensure that L will remain less than 0.95 during core alterations and fuel handling operations.
Because no core alterations or fuel handling operations are planned during the current Unit 2 forced outage, and because, as was shown above, the proposed 1900 ppm boron concentration limit will ensure that remains less than 0.95 under all postulated conditions, Northern States Power concludes there is reasonable assurance that the health and safety of the public will not be adversely affected by the proposed Technical Specification change.
Determination of Slanificant Hazards Considerations The proposed changes to the Operating License have been evaluated to determine whether they constitute a significant hazards consideration as required by 10 CFR 50, Section 50.91 using the standards provided in Section 50.92. This analysis is provided below:
- 1. The proposed amendment will not involve a sianificant increase in the probability or conseouences of an accident previously evaluated.
As the proposed changes would only be in place for a limited duration and because they do not involve any physical changes to the reactor coolant system and no equipment is operated in a new or different manner, the proposed
I I.
Exhibit A y P ge 6 cf 8 l
changes do not involve an increase in the probability of an accident previously evaluated.
Because the proposed 1900 ppm boron concentration limit will ensure that K.n remains less than 0.95 under all postulated conditions, this change will not
!mpact the consequences of any accident previously evaluated.
Therefore, based on the conclusions of the above analysis, the proposed changes will not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. The proposed amendment will not create the possibility of a new or different kind of accident from any accident oreviously analyzed.
The proposed amendment would not create the possibility of a new or different kind of accident previously evaluated, as it only involves a limited duration reduction in the reactor system boron requirements for Mode 6 operation. There is no change in plant configuration, equipment or equipment design. No equipment is op6 rated in a new manner. Thus the changes will not create any new or different accident causal mechanisms. The accident analysis in the Updated Safety Analysis Report remains bounding.
Therefore, based on the conclusions of the above analysis, the proposed changes will not create the possibility of a new or different kind of acc dent.
- 3. The proposed amendment will not involve a sianificant reduction in the maroin of safety.
The required Mode 6 boron concentration ensures that Kon will remain less than 0.95 so that adequate margin to criticality is always maintained during core alterations and fuel handling operations. No core alterations or fuel handling will be necessary and none are planned during the operations to ba performed in Mode 6 for resolution of the Unit 2 CRDM leakage. With the exception of reactor vessel head removal, the reactor configuration will be the same as it is during Mode 5, where K.n is only required to be less than 0.99. Because no core alterations or fuel handling operations are planned during the current Unit 2 outage, and because the proposed 1900 ppm boron concentration limit will ensure that X.n remains less than 0.95 under all postulated conditions, there is no significant reduction in the plant's margin of safety, it has also been shown that the proposed 1900 ppm Mode 6 boron concentiation will ensure that the analysis of a dilution accident as described in the lipdated Safety Analysis Report remains bounding.
Based on these factors, the changes proposed by this License Amendment Request will not result in a significant reduction in the plant's margin of safety.
7 Exhibit A Pega 7 of 8 Conclusion m
Based on the evaluation above, and pursuant to 10 CFR 50, Section 50.91, Northern States Powar Company has determined that operation of the Prairie Island Nuclear Generating Plant in accoroance with the proposed license amendment request does not involve any significant hazards considerations as defined by NRC regulations in 10 t CFR 50, Section 50.92.
. Environmental Assessment
[ Northern States Power has evaluated the proposed changes and determined that:
$ 1. The changes do not involve a significant hazards consideration,
- 2. The changes ao not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or k
- 3. The changes do not involve a significant increase in individual or cumulative L occupationai radiation exposure.
Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR Part 51 Section 51.22(c)(9). Therefore, pursuant to 10 CFR Part 51 Section 51.22(b), an environmental assessment of the proposed changes is not required.
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Exhibit A P:ge 8 cf 8 Attachment Prairie Island RCCA Locations 1 2 3 4 5 6 7 8 9 10- 11 12 13 A A B RCCA RCLA B C RCCA RCCA RCCA C D RCCA RCCA D E RCCA RCCA E RCCA .
F RCCA RCCA ;'CCA RCCA F G RCCA RCCA RCCA G R RC H- RCCA RCCA RCCA RCCA H 1 RCCA RCCA I RC J RCCA RCCA J
% RCCA RCr5 RCCA K L RCCA RCCA .L M M i 2 3 4 5 6 7 8 9 10 11 12 13 1
_ _ _ _ _ _ .