ML20246D923
| ML20246D923 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 08/18/1989 |
| From: | Heitner K Office of Nuclear Reactor Regulation |
| To: | Weiss S Office of Nuclear Reactor Regulation |
| References | |
| TAC-73124, NUDOCS 8908280327 | |
| Download: ML20246D923 (35) | |
Text
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- M [NITED t STATES
$" j[= 3 e.( k, ' p, ' NUCLEAR REGULATORY COMMISSION g; .4 ' j WASHINGTON, D. C. 20$$$
. August 18, 1989
'% Q Docket No. 50-267
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MEMORANDUM FOR: 'Seymour H. Weiss, Director Non-Power Reactor, Decommissioning and Environmental Project Directorate Division of Reactor Projects - Ill, IV, Y and Special Projects FROM: Kenneth L. Heitner, Project Manager Non-Power Reactor, Decommissioning and Environmental Project Directorate Division of Reactor Projects - III, IV, Y and Special Projects
SUBJECT:
SUMMARY
OF MEETING WITH PUBLIC SERVICE COMPAhY OF COLORADO (PSC) TO DISCUSS DEFUELING 0F FORT ST. VRAIN (FSV) JULY 18,1989(TACNO.73124)
This meetina was requested by PSC to further discuss issues related to the cefueling cf FSV. The attendees at this meeting are listed in Enclosure 1.
Material supplied by PSC at this meeting is Enclosure 2.
Background
This meeting was a followup to letters dated May 11 and May 15, 1989 submitted to the staff by PSC. In these letters PSC indicated two fur.damental changes in its approach to the defueling of FSV and'its conduct of operatior.s during the oefueling period. Tre first change is that PSC would conduct the defueling of FSV under 10 CFR 50.59. The second is that PSC would not request further Techn1 cal Specification changes specifically for defueling, or as part of the Technical Specification Upgrade Program (TSUP). These viewpoints were presented by PSC at a meeting with NRC managers on May 11, 1989.
PSC's Restatement of Position PSC restated their position that NRC approval was not required for defueling
- of Fort St. Vrain. Key points in PSC's presentation were that:
- There were no unreviewed safety questions,
- There would be an increased safety margin during defueling, and
- That defueling safety concerns were adequately controlled by existing Technical Specification requirements.
In their discussion of the proposed defueling, PSC noted the importance of
-reir.taining the core configuration to be a right circular cylinder. This
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CONTACT:
K. Heitner, NRR/PDNP l- 492-1333 .
8908280327 890818 7 i ADOCK 0500 gDR
Seymour H. Weiss geometry was capable of being analyzed by accepted methods. PSC also noted their interpretation of Interim Limiting Condition for Operation (LCO) 3.1.6 as being only applicable to the portion of the core containing fuel. The control rods would be withdrawn in regions where dummy fuel elements were present. PSC admitted the proposed dummy fuel elements were not discussed (or described) in Section 6.0 of the FSV Technical Specifications (TS). However, PSC stated that the safety basis were the shutdown margin requirements in other portions of the TS. PSC also noted that some uncertainty still existed about the ability of the excore startup neutron detectors to monitor the defueling process.
Potentially, the count rate for detectors could fall below TS allowable values before the core was suberitical with all rods out. PSC was considering whether a license amendment would be needed later in the defueling process. PSC also stated that they had not decided whether to reevaluate control rod worths prior to starting defueling.
Staff's Statement of Concerns The staff presented its concerns with respect to the proposed method of defueling.
These concerns would have to be adequately addressed in a licensee 50.59 evaluation. The staff had considered a number of other documents, including PSC letters dated January 20 and June 16, 1989 and the summary of the previous meeting with PSC on defueling dated March 13, 1989 in identifying these issues.
First, the proposals for defueling the reactor are significantly different from the original reactor fueling as defined in the FSV Final Safety Analysis Report (FSAR)(Section13.3). The fueling was done by layers with temporary absorber strings. The defueling is proposed to be by core regions with a radially inward pattern. Hence, the defueling would differ significantly from previously described fueling and refueling activities.
The second issue of concern was the ability of the startup detectors to provide adequate measurement and monitoring of the core's subcritical con-figuration curing the defueling process. Again, the proposed use of the startup detectors is different from the approach used during reactor fueling where detecters were placed in the reactor core. The changes in core geometry during defueling are significant compared to those involved with a normal refueling. The capability of the startup channels to detect criticality would
. beaffectedbythereducedproductionofneutronspergourceneutron. The adequacy of the source range trip, currently set at 10 counts per recond (cps), could also be significantly affected. These questions are unique to the defueling scerario. The purpose of the startup detectors and the associated trip setpoint is to scram the reactor should an inadvertent criticality occur during the defueling operation. (FSC has also noted the problem of maintaining an adequate count rate in the detectors.)
The third issue relates to the description of the Reactor Core - Design Features. Section 6.0 of the Fort St. Vrain FSAR addresses design features, in particular, Section 6.1 addresses the design features of the reactor core.
The objective of this section states: "to define vital design characteristics of the reactor core to control changes in the design features." In the 1
1
. _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ ___________.___.__.__m
b Seymour H. Weiss f:.-
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I4 discussions on March 7,'1989 '.concerning FSV defueling, PSC proposed the use of I, dumy fuel blocks containing boron material. As the defueling process proceeded,.
L .the absence of fuel'and-its replacement with the dummy fuel blocks would play.
'an increasing and significant role in maintaining the. core subcritical (i.e.,
f.
.providing reactivity control). Once a core region is defueled, the dumy fuel blocks provide the negative reactivity. In addition, the dumy fuel blocks maintain the' structural integrity of_ the core. However, the dummy fuel blocks
-are not described in TS.Section 6.1 - Reactor Core - Design Features.. The materials of construction are changed from those specified.
Finally, by letter dated July 10, 1985,- PSC comitted to' operating FSV under the-Interim Technical Specifications for Reactivity Control (attached to that letter). A literal reading of Certain Interim TS are potentially inconsistent
, with PSC's proposed defueling approach. For example, Interim TS 3.1.6. A.1 only allows up to two control rod pairs to be withdrawn for refueling. PSC's s proposed defueling would require more than two control rod pairs withdrawn.
The 50.59 evaluation associated with defueling would have to clarify the intent of the interim TS for this proposed activity and to show that the comitment.
to operate under the interim TS is not changed in a manner that would require NRC review and approval.
Conclusions No conclusions were reached relative to staff agreement with PSC defueling FSV' under the terms of 10 CFR 50.59. The staff stated that the decision to proceed with evaluating the defueling under 10 CFR 50.59 remained with PSC. However, PSC in their 10 CFR 50.59 analysis will have to address the issues the staff ']
identified including how the requirements of 10 CFR 50.59 have been met. The "l staff requested that PSC provide a copy of the 'defueling Safety Analysis Report !
(SAR) when it was complete in any event. PSC agreed to this request. PSC also agreed to provide the SAR for plant coastdown past 300 equivalent full power days. 1 Technical Specification changes under the Technical Specification Upgrade Program were not discussed at this meeting.
lS l Kenneth L. Heitner, Project Manager Non-Power Reactor, Decommissioning and Environmental Project Directorate Division of Reactor Projects - III, IV, V and Special Projects
Enclosures:
As stated ()(-o[
cc w/ enclosures:
See next page (
DISTRIBU1 ION
$DocketJ11e# NRC PDR Local PDR J. Sniezek PD4 Reading F. Hebdon K. Heitner OGC-Rockville J. Sharkey, EDO F. Litton, EMTB J. Miller, TSB E. Jordan B. Grimes ACRS (10) L. Kopp G. Holahan T. Westerman, RIV J. Partlow E. Tomlinson T. Martin (RegionIV)
DOCUMENT NAME: FSV 8980 y
- SEE PREVIOUS CONCURRENCES: KHeitner for: r, y p PD4/LA* PD4/PM* PD4/D* G'A)ADR4 PNoonan KHeitner:sr FHebdon GHolahan 6 07/2C/89 07/20/89 07/20/89 08/jr/89
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,,, Seymour.H. Weiss. discussions on March 7, 1989 concerning FSV defueling, PSC proposed the use of dumy fuel blocks containing boron material. As the defueling process proceeded, I the absence of fuel and its replacement with the dummy fuel blocks would play an increasing and significant role in maintaining the core suberitical (i.e.,
providing reactivity control). Once a core region is defueled, the dumy fuel blocks provide the negative reactivity. In addition, the dumy fuel blocks wintain the structural integrity of the core. However, the dummy fuel blocks sre not described in TS Section 6.1 - Reactor Core - Design Features. The materials of construction are changed from those specified.
Finally, by letter dated July 10, 1985, PSC comitted to operating FSV under the Interim Technical Specifications for Reactivity Control (attached to that letter). A literal reading of Certain Interim TS are potentially inconsistent with PSC's proposed defueling approach. For example, Interim TS 3.1.6.A.1 only allows up to two control rod pairs to be withdrawn for refueling. PSC's proposed defueling would require more than two control rod pairs withdrawn.
The 50.59 evaluation associated with defueling would have to clarify the intent of the interim TS for this proposed activity and to show that the comitment to operate under the interim TS is not changed in a manner that would require NRC review and approval.
Conclusions No conclusions were reached relative to staff agreement with PSC defueling FSV under the terms of 10 CFR 50.59. The staff stated that the decision to proceed-with evaluating the defueling under 10 CFR 50.59 remained with PSC. However, PSC in their 10 CFR 50.59 analysis will have to address the issues the staff identified including how the requirements of 10 CFR 50.59 have been met. The staff requested that PSC provide a copy of the defueling Safety Analysis Report (SAR) when it was complete in any event. PSC agreed.to this request. PSC also agreed to provide the SAR for plant coastdown past 300 equivalent full power days.
Technical Specification changes under the Technical Specification Upgrade Program were not discussed at this meeting.
Kenneth L. Heitner, Project Manager .
Non-Power Reactor, Decommissioning i and Environmental Project Directorate
. Division of Reactor Projects - III, IV, V and Special Projects
Enclosures:
. As stated cc w/ enclosures:
See next page DISTRIBUTION Docket File NRC PDR Local PDR J. Sniezek PD4 Reading F. Hebdon K. Heitner OGC-Rockville J. Sharkey, EDO F. Litton, EMTB J. Miller, TSB E. Jordan l B. Grimes ACRS (10) L. Kopp G. Holahan T. Westerman, RIV J. Partlow E. Tomlinson T. Martin (Region IV)
DOCUMENT NAME: FSV 8980 ,y
- SEE PREVIOUS CONCURRENCES: KHeitner for: f p PD4/LA* PD4/PM* PD4/D* /S'AIADR4 PNoonan KHeitner:sr FHebdon GHolahan !
07/20/89 07/20/89 07/20/89 08/lt/89 !
l L_ - _ - _ _ -
\
Seymour H. Weiss .
discussions on March 7, 1989 concerning FSV defueling, PSC proposed the use of dumy fuel blocks, containing boron material. As the defueling process proceeded, the absence of fuel end its replacement with the dummy fuel blocks would play an increasing and significant role in maintaining the core subtritical (i.e.,
providing reactivity control). Once a core region is defueled, the dummy fuel blocks provide the negative reactivity. In addition, the dummy fuel blocks maintain the structural integrity of the core. However, the dummy fuel blocks are not described in TS Section 6.1 - Reactor Core - Design Features. The materials of construction are changed from those specified.
Finally, by letter dated July 10, 1985, PSC comitted to operating FSV under the Interim Technical Specifications for Reactivity Control (attached to that letter). A literal reading of Certain Interim TS are potentially inconsistent with PSC's proposed defueling approach. For example, Interim TS 3.1.6.A.1 only allows up to two control rc,d pairs to be withdrawn for refueling. PSC's proposed defueling would require more than two cont' 01 rod pairs withdrawn.
The 50.59 evaluation associated with defueling wouN have to clarify the intent of the interim TS for this proposed activity and to show that the comitment to operate under the interim TS is not changed in a manner that would require NRC review and approval.
Conclusions No conclusions were reached relative to staff agreement with PSC defuelino FSV under the terms of 10 CFR 50.59. The staff stated that the decision to proceed with evaluating the defueling under 10 CFR 50.59 remained with PSC. However, PSC in their 10 CFR 50.59 analysis will have to address the issues the staff identified including how the requirements of 10 CFR 50.59 have been met. The staff requested that PSC provide a copy of the defueling Safety Analysis Report (SAR) when it was complete in any event. PSC agreed to this request. PSC also agreed to provide the SAR for plant coastdown past 300 equivalent full power days.
Technical Specification changes under the Technical Specification Uperade Program were not discussed at this meeting.
Kenneth L. Heitner, Project Manager Non-Power Reactor, Decommissioning l and Environmental Project Directorate l Division of Reactor Projects - III, IV, V and Special Projects
Enclosures:
As stated cc w/ enclosures:
See next page
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Fort St. Vrain 4
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Mr. D. W. Warembourg, Manager Robert M. Quillen, Director Nuclear Engineering Division Radiation Control Division Public Service Company Department of Health of Colorado 4210 East lith Avenue P. O. Box 840 Denver, Colorado 80220 Denver, Colorado' 80201-0843
. Mr. David Alberstein, Manager Mr. Charles H. Fuller Fort St. Vrain Services Manager, Nuclear Production GA International Services Corporation and Station Manager i Post Office Box 85608 Public Service Company of Colorado San Diego, California 92138 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Mr. H. L. Brey,. Manager .
Nuclear Licensing and Resource lir. P. F. Tomlinson, Manager Management Division Quality Assurance Division Public Service Company of Colorado Public Service Company of Colorado P. 0. Box'Ba0 16805 Weld County Road 19-1/2 Lenver,. Colorado 80201-0640 Platteville, Colorado 80651 Senior Resident inspector Mr. D. ' D. Hock U.S. Nuclear Regulatory Comission President and Chief Executive Officer P. O. Box 640 Public Service Company of Colorade Platteville, Colorado 80651 Post Office Box 640 Denver, Coloraco 80201-0840 Kelley, Standfield & 0'Donnell ATTN: Mr. J. K. Tarpey Commitment Control Program Public Service Company Building Coordinator Room 900 Public Service Company of Colorado 550 15th Street 2420 W. 26th Ave. Suite 100-D Denver, Colorado 80202 Denver, Colorade 80211 Regional Administrator, Region IV Mr. R. O. Williams, Jr.
U.S. Nuclear Regulatory Commission Senior Vice President, Nuclear Operations 611 Ryan Plaza Drive, Suite 1000 Fublic Service Company of Colorado Arlington, Texas 76011 Post Office Box B40 Denver, Colorado 80201-0840 Chairman, Board of County Commissioners of Weld County, Colorado Greeley, Colorado 80631 j Regional Representative Radiation Programs Environmental Protection Agency 1 Denver Place 999 18th Street, Suite 1300 Denver, Colorado 80202-2413 l
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- f. ; Enclosure l' 4
je 'Attenoees at NRC-PSC Meeting of July 18, 1989
- Name' Organization Ken L; Heitner: NRC/HRR/PD-IV
- Don Warenbourg - PSC A. Clegg Crawford - PSC Charles H.. Fuller PSC H.'L. Brey. PSC R. H. Vollmer PSC/Tenera D.'. Alberstein .. . General Atomics Dania11e Weaver Nucleomes Week Larry Kopp NRC/NRR/SRXB Gary Holahan ' NRC/NRR/DRSP Tom Westerman NRC/RIV Ja:nes Partlow NRC/NRR Ed Tomlir. son . NRC/NRR/PD-IV 0
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Enclosure 2
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AGENDA NRC PSC EXECUTIVE MEETING JULY 18.1989
- 1. INTRODUCTION / OBJECTIVE 'Crawford !
s
- 11. DEFUELING STATUS Warembourg 111.' DEFUELING SAR
SUMMARY
Fuller IV. CONCLUSIONS Crawford V. PROPOSED ACTIONS Crawford VI. DISCUSSION l
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l INTRODUCTION L MEETING OBJECTIVE
- CLARIFY THE NEED FOR NRC APPROVALS, IF ANY, TO DEFUEL FORT ST. VRAIN
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L INTRODUCTION (CONTINUED)
PROPOSED ACTIONS SUBMIT FINAL DEFUELING SAR FOR NRC INFORMATION
= Will be submitted with a cover letter including:
a Brief description of defueling process 4 Assumptions and basis for 50.59 conclusions SET UP A JOINT WORKING MEETING TO RESOLVE REMAINING TECHNICAL ISSUES
.s ..
.? ,4 1 PSC POSITION-
- NRC APPROVAL NOT REQUIRED FOR DEFUELING
= Not required by regulations 4 a Not required for typical LWR defueling a :No defueling unreviewed safety questions
=- Increased safety margin during defueling i Less reactive 4- Fission products decaying 4 Reduced heat generation
. )
-a No immediate threat to public heal.th or safety a
Defueling safety concerns adequately controlled by existing i Technical Specification requirements
+ -
a.
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DEFUELING OVERVIEW
- Defueling milestones
= Be ready to begin defueling as early as November 1,1989 s Base case defueling plans begin defueling January 2,1990 m
Optimum case begin defueling July 1,1990
- Strategy
= Keep defueling similar to refueling
= Defuel by region a
Defuel by ring outer to inner to maintain validity of computer models a
Replace fuel elements with boronated, HLM graphite defuel elements a Utilize existing Technical Specifications ,
- SAR for coast down
- SAR for defueling
m 1' '. ..
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L SAFETY ANALYSIS REPORT
.FOR REACTOR DEFUELING l
CONTENTS l.
- 1. INTRODUCTION AND
SUMMARY
- 2. DEFUELING GENERAL DESCRIPTION 2.1. Defueling Method 2.2 Defueling Element Design 2.3 Lumped Poison Pin Design
- 3. NUCLEAR ANALYSIS 3.1 Neutron Sources and Reactivity Monitoring 3.2 Shutdown Margin During Defueling 3.3 Shutdown Margin Verification . .
3.4 Effects of Further Depletion on Shutdown Margin
- 4. THERMAL-HYDRAULIC AND MECHANICAL ANALYSIS >
4.1 Thermal-Hydraulic Performance During Defueling
, 4.2 Mechanical Performance
- 5. SAFETY ANALYSIS 5.1 Introduction 5.2 Events Requiring Further Evaluation 1
. - = - - - - - - - - - -
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[:j 5.31 Events No. Longer Credible F..'
r- J 5.41 Conclusions e - 6. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS
,z ' 7. REFERENCES p
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.I DEFUELING SEQUENCE OBJECTIVE
- Utilize a shrinking core concept to' ensure a core geometry consistent with established Fort St. Vrain core physics analysis models
- Ensure sufficient shutdown margin at all points in the-sequence J
- Ensure a neutron count rate on the startup channels
~t hat is adequate for monitoring core reactivity until such monitoring is no longer needed
- Minimize the number of fuel handling machine movements
- Provide for efficient fuel deck logistics I
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'A CORE PREPARATIONS
- Remove metal clad (boronated) top reflector Regions 3, 6,10, and *. Insta' ll neutron sources in Regions'3 and 6
-* Replace metal clad top reflector in Regions 3,6,10, and 16 with non-boronated elements 1
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DEFUELING SEQUENCE
- Defuel by ring - outer to inner
- Shrinking l ' core concept maintaining right circular cylinder configuration
- Regions 3 and 6'which contain neutron sources are the last regions to be defueled I l i
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IC INITIAL CORE scoLENCE rwacR h 8,9 0
SEGMENT 8 OR 9 DEFUELING Figure 2-1 Reactor Core Defueling Sequence 2-10
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DEFUELING ELEMENTS
- HLM equivalent graphite elements
- a Satisfies all reactor physics, thermal and overall environmental requirements a Similar in structural strength to H-327 or H-451 and conservatively meets or exceeds all therrnal hydraulic requirements a HLM graphite is currently used for permanent side reflector elements a Boron' is presently installed in the side
- reflectors
- Boron carbide lumped poison pins m Equivalent of 100 ppm is adequate for reactivity control during defueling a Poison loading provides for the equivalent of 350 ppm for conservatism l
l L___---__------- - -
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m A region filled:with defueling elements is
.: at least equivalent to the control rod worth l
- Overall' design equivalent to existing fuel elements with the exception of inner ring of coolant holes and use of blind holes for lumped poison pins
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LUMPED POISON PINS Figure 2-2 Defueling Element Top View (Dowels Not Shown) -l.
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REACTIVITY MONITORING -
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- ' Accomplished using the existing stadup channels Count rate will be further enhanced by inserting additional neutron sourcea in Regions 6 and 3 Before proceeding with the defueling of a region, a shutdown margin confirmation test is done in accordance with the current Technical Specifications.
Count rates are monitored during this test Count rates are recorded before and after a region is defte ed The current monitoring requirement and shutdown margin assessment testing is relaxed when:
The calculated Keff of the remaining fuel is less than or equal to 0.95 with all rods withdrawn Physical demonstration of suberiticality is performed by withdrawing all control rods and verifying subcriticality. This would be the final shutdown margin verification test
., - ?
- _.,.
SHUTDOWN MARGIN ASSUMPTIONS i
- Uses the'before-mentioned sequence
= Approximately 0.5% per 50 EFPD burnup
- Two rods in the sequence are withdrawn. Calculations
.also assumed that the Region 1 Rod is withdrawn.
Current intention is to not withdraw the Region 1 Rod.
- Use 12 LPP/ block design using blind coolant holes.
m Minimum diameter a Minimum stack height '
1
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- Minimum concentration l
- Calculations performed with gauge code model 1
l
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e FINAL SHUTDOWN MARGIN VERIFICATION The criterion is calculated Keff Isss than or equal to 0.95 with all rods out
- With increasing core burnups, the ci terion 'is met with more regions of fuel remaining Shutdown is after steady state operation at 82% pcwer CONCLUSION:
At 200 EFPD, the criteria is met with 8 regions of fuel left (Regions 1 through 7, and 16)
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' ACCIDENT EVALUATION
- Accidents involving depressurization are not credible since the PCRV will be maintained near atmospheric LOFC with one PCRV liner cooling loop was analyzed.
The cooling loop was staded within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the LOFC which occurred 100 days after shutdown RESULTS:
a All core internal temperatures are lower than the temperatures- experienced during normal operation Seismic event was analyzed with a region of fuel removed RESULTS:
A fallen element may break, fuel particle integrity remains intact, and fuel temperatures stay below 2.900 degrees
- Core support block would be damaged only if struck in the center. However, core suppod posts are not damaged and the overall core support structural integrity is maintained i
.
- No chance for recriticality i
- No safety consequences but PSC must deal with a j different cleanup problem j i
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l 1-ACCIDENT EVALUATION (CONTINUED)
= Inadvertent criticality accident. A postulation was made that during.the shutdown margin assessment test, the wrong rod was accidentally pulled. This rod was assumed to be'the maximum worth rod.
RESULTS:
= No criticality was calculated a However for three regions. Keff's of about 0.99 were calculated (at 155 EFPD)
= The number of regions that cause this problem decrease with core burnup ACTION:
a Enhance core safety by providing total assurance that there will be no power to these rod drives, thus making the accident incredible m_______._.m _ _ _ _ . _ _ _ _ _ _ _ _ _
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ACCIDENT ANALYSES l 1
= Earthquakes a Reactivity accidents 4 Excessive removal of poison A Loss of fission product poison a Core rearrangement a introduction of steam to core 4 Sudden decrease reactor temperature a Rod withdrawal accidents a Column deflection / misalignment
-a Misplaced fuel element a Coolant channel blockage e incidents involving electricai system a Loss of cooling a Leaks inside primary coolant system i Steam generators 4 Moisture ingress a Fuel storage accidents n ----___-----,_---__--------a-----------------__-___---____n- - - - - - - - - _ - _ _ - - - - - - - - . - - - - - - - - - . - - - - - - - - - - - - - - - - . _ _ _ _ . _ - - - - - - - - - _ _ _ _ - - - - - _ _ - - - _ _ _ _ _ _ _ _ _ _ . - - - _ - - - _ _ _ -
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OVERALL SAR CONCLUSIONS a No changes to the existing or interim Technical Specifications are required to proceed with defueling
=
- There are no unreviewed _ safet) or environmental issues e A Technical Specification change may be required toward the end of defueling concerning startup channe: count rate and verification of Keff of less than 0.95 with all rods withdrawn l
l a -- --- - - - - - - - _ -
n ,.: . .
e CONSERVATISM e An uncertainty error of 0.01 Delta K was assumed in a well understood core a Remain shutdown by at least 0.01 Delta K even if a
- rod withdrawal accident is experienced with highest worth rod a Except for Region 33 at 200 EFPD 4 For all cases after 300 EFPD
= As core burnup continues the SAR calculations of shutdown margin at 155 EFPD become even more
. conservative
, s' SAR calculations assume Region 1 Rod is withdrawn
. c COMPARISCN i k!
l .
I I l REFUELING l DEFUELING l l l l l l l l PRELIMINARY SEQUENCE IS SELECTED ISAME I l(EDTH INNER AND OUTER REGIONS) l(OUTER) l I I I I I I lADE00ACY OF SHUTDOWN MARGIN IS ISAME l l CONFIRMED BY CODE I 1 i i l i i l ISHUTDOWN MARGIN VERIFICATION RCDS lSAME 1 l ARE SELEC'iED l l l(0.DIaK + NEW FUEL ap + TEMPERATURE l(0.01aK + TEMPERATURE I l CEFECT, LCO 3.1.6) l DEFECT) l l I I I I I ISAR IS-VRITTEN UNDER EP.59 AS BASIS ISAME l lFCR REFUELING 1 1 I l l l l 1 l CHANGES TO SEQUENCE OR SHUTDOWN MARGIN ISAME l VERIFICATION-R005 ARE RE-EVALUATED l l lUNDER E0.E9 l l . ,
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INEW FUEL IS INSERTED IBORONATED DEFUELING l l(POSITIVE REACTIVITY) lELEMENTS (NEGATIVE l 3 g IREACTIVITY) g l i i i l
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l DEFUELING SEQUENCE CONTROL
= PSC has demonstrated that 50.59 adequately controls the sequence and changes to the sequence a Interim Technical. Specification LCO 3.1.4 already provides the' appropriate .iutdown margin requirements
= Interim Technical Specification Surveillance SR 4.1.6 already requires an analysis and verification test of the shutdown margin a The requirement to have no more than two control rods simultaneously removed (loterim Specification LCO 3.1.6) is understood by PSC to apply to regions with fuel
. .r.
DESIGN FEATURES I
= Defueling elements are not explicitly discussed in Section 6, however, use of boronated elements is discussed
- a. Purpose of' Section is to describe " Design Characteristics of Special importance to Each of the Physical Barriers and to the Maintenance of Safety Margins Which Have Not Been Covered in Any Other Specifications" a The ' safety margin' associated with defueling elements is the 0.01 Delta K shutdown margin, which is already addressed in current Technical Specifications L_ __ __
.. :1 l.
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1 FUEL HANDLING / FUEL STORAGE TECHNICAL SPECIFICATIONS
= Existing Technical Specifications address fuel handling and fuel. storage requirements e A comparison of existing and TSUP Specifications does not identify any significant differences
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CONCLUSIOPLS
- WE HAVE NO UNREVIEWED SAFETY ISSUES. FORT ST. VRAIN CAN BE SAFELY DEFUELED USING EXISTING TECHNICAL SPECIFICATIONS. THEREFORE, P9C HAS THE AUTHORITY UNDER 10 CFR 50.59 TO PROCEED .VITH DEFUELING FORT ST.
VRAIN
- NRC DEFUELING APPROVALS ARE, IN THE OPINION OF PSC, NOT REQUIRED.
-_-_-__ ._-.-..._._____.__ _______--_- - __ -__._ a
v . . -.; . 3 .
e j ._ ' o-PROPOSED ACTIONS
- SUBMIT FINAL DEFUELING SAR FOR NRC INFORMATION SET UP A JOINT WORKING MEETING TO RESOLVE REMAINING
~ TECHNICAL ISSUES f
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