ML19325D422
| ML19325D422 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 10/17/1989 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML19325D421 | List: |
| References | |
| NUDOCS 8910240025 | |
| Download: ML19325D422 (4) | |
Text
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WA$HING TON. D. C. 20bbb i.
FRELIMINARY SAFETY EVALUATION AND REQUEST FOR ADDITIONAL INFORMATION CONCERNING DEFUELING TORT ST. VRAIN NUCLEAR GENERATING STATION PUBLIC SERVICE COMPANY OF COLORADO DOCKET NUMBER 50-267
1.0 INTRODUCTION AND BACKGROUND
By letter dated December 8,1989 the licensee tnnounced the proposed early terminatioaofFortSt.Vrain's(FSV)nuclearoperations,andinitiationof plans for an orderly defueling. The licensee held a meeting with the staff on j
March 7, 1969 to discuss defueling. By letters dated May la, and 15, 1909, the licensee stated it would conduct the defueling of FSV under 10 CFR 50.59. A second meeting was held with the staff on July 18, 1969 to discuss defueling.
l In these meeting minutes dated August 18, 1989 the staff stated that although l
it was the licensee's decision relative to proceding with defueling under i
50.59, the safety analysis will have to address the issues the staff identified including how the requiresents of 10 CFR 50.59 have been met. By letter dated August 16, 1989, the licensee submitted it's Safety Analysis Report (SAR). The licensee concluded that no unreviewed safety questions existed that would preclude proceeding with the defueling process. However, they have not provided i
assurance that they have considered and resolved the issues identified by the
- staff, j
By letters dated September 14, 1989, the licensee submitted certain Technical i
specification (TS)changestosupporttheproposeddefueling.
[
informationconcerningdefuelingfromtheIIcensee.affrequestedadditional Additionally, by letter dated July 25,1989 the st The licensee responded to these questions by letter dated August 24, 1989. This safety evaluation (SE) is a preliminary review of the licensee's submittals and the above referenced sieeting sunraries. This SE also contains requests for additional information i
concerning the proposed FSV defueling.
2.0 EVALUATION t
The etaluation covers four areas of staff concerns. These are:
Reactivity Evaluation, Reactivity Monitoring, Accident Analysis, and Redundancy of Reactivity Control. Technical Specification changes associated with the defueling are discussed in Section 3.0.
i 2.1 Reactivity Evaluation The staff has reviewed the licensee's submittal concerning the boronation model dated August 24,1989(Question 1). The licensee and staff agree that B910240025 891017 gDR ADOCK OS ] g7
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2 the bortnation model should give the sare results for any level of boronation iri the case of zero regions defueled.
It is the staff's understanding that the results of these calculations are de)endent on sequence of regions unrodded and defueled. The sequence chosen for tic SAR is different than that presented to I
the staff at the March 7, 1989 meeting.
I The licensee should provide actual calculated resu'Its with the boronation mcdel for 0 ppm boron for defueling of the first 10 regions. The rod withdrawal and l
defueling sequence should be consistent with that presented in the defueling SAR Table 3-2.
The licensee should demonstrate that these calculations are consistent with the 10 pin boron results presented in the SAR.
The staff's position is that censistent results from the model will resolve this concern.
Cycle 4 Calculations f
Table 5-10 of the Cycle 4 SAR gives the shutdown margin at 80'F with complete Pa-233 decay as 0.096aK at 150 EFFD and 0.114 AK at 300 EFPD.
Table 3-2 of the defueling SAR gives K eff of 0.9183 (shutdown margin *.0817) at 155 EFPD for the same conditions. The shutdown margin for 300 EFPD =.0986.
It is the l
staff's understanding that these differences arise from minor differences in the models and model parameters usted in these calculations. The licensee should provide a brief explanation of these differences.
2.2 Reactivity Monitoring The licensee has proposed the use of the startup detectors to provide measurement and monitoring of the core's subtritical configuration during the defueling i
process.
The proposed use of the startup detectors is different from the i
approach used during reactor fueling where detectors were placed in the reactor i
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 be affected by the reduced production of neutronsgersourceneutron. The adequacy of the source range trip, currently set at 10 countspersecond(cas) could also be significantly affected.
Thesequestionsareuniquetotielicensee'sproposeddefuelingscenario. The purpose of the startup detectors and the associated trip set point is to alert the operators and potentially scram the reactor should an inadvertent criticality occur during the defueling cper: tion.
The licensee's analysis of rea:tivity nonitoring does not provide a quantitative evaluation of the projected count rate at the start-up channel detectors. A normal refueling operation only involves one refueling region at a time.
Experience has shown that an adequate count rate can be maintained under thesc conditions. In defueling, changes are made to many regions with no assurance that an adequate count rate can be maintained. Thus, the staff concludes the licensee has not provided an adequate basis for determining the adequacy of the proposed reactivity monitoring scheme.
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The licensee should provide a best estimate arojection of the startup destectors count rate vs regions defueled. Tie effect of modifications proposed in the defuelir.g SAR including modifications of boronated plenum elements should also be quantitativcly evaluated.
Alternatively, the licensee should provide a specific proposal for a different reactivity monitoring scheme.
(This could include the use of temporary incore monitors as appropriate).
,2,3 Accident Analysis The licensee's defueling SAR, Section 5.22 argues that "A Rod Pair Withdrawal 1.ccident During Startup Operations at Sourck Levels or at Very Low Power" is r.ot credible because the core will be shutdown at defueling conditions.
However, the licensee's own calculations in the defueling SAR, Table 3-2 show the reactor is critical with onfy four or five control rods withdrawn.
ThelicenseestatesthatcertaincoNrolroddriveswillbedeenergizedto prevent an accidental criticality, litwever', the possibility of such accidents is not precluded, especially during.thutdown margin testing. This accident would be similar to that d4 :r % :1.r FSAR Section 14.2.2.7.
The licensee should present a note complete discussion of this issue. The goal of the dircussion should demonstrate that appropriate safety limits are r.ot exceeded oc are bounded by the existing FSAR analysis.
2.4 Redundancy of Reactivity Control The staff has reviewed the licensee's response submitted August 24, 1989 concerning redundancy of reactivity control. The licensee has demonstrated in the defueling SAR that a combination of the control rods and boronated duuy blocks will maintain the reactor suberitical throughout the defueling sequence. However, no comparable calculations have been performed for the reserve shutdown system (RSS). The RSS is the independent means of reactivity
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control needed to satisfy FSAR Design Criteria 27.
FSAR Section 3.5.3.3 provides an acceptable method of demonstratirg the capability of the RSS to independently shutdown the reactor. The licensee should provide an equivalent analysis for the proposed defueling sequence.
3.0 TECHNICAL SPECIFICATIONS This section discussed TS changes needed to support the proposed defueling
- plan, 3.1 Reactor Core-Design Features Section 6.0 of the Fort St. Vrain TS 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 l
reactcr core to control changes in the design features." In the defueling l
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i SAR the licensee proposed the use of dunny fuel blocks containing boron L
saterial. As the defueling process proceeded the dunuy fuel blocks would j
playanincreasingandsignificantroleinmaIntainingthecoresuberitical l
(i.e., providing reactivity control). Once a core region is defueled, the dunay-fuel blocks provide the negative reactivity in place of the control rods and the RSS. The dupey fuel blocks are not described in TS Section 6.1 -
l Reactor Core - Design Features. The materials of construction are also changed i
from the H-327 and H-451 graphites specified.
It is the staff's position that this TS is applicable at all times and certainly when there is fuel in the reactor.
In view of the above, the staff concludes that the proposed duasqy fuel blocks described by the licensee cannot be placed in the reactor core without a j
suitable amendment to TS Section 6.1.
A proposed TS change to Section 6.1 was received from the licensee by letter dated September 14, 1989.
l 3.2 Reactivity Control TS The current Interim TS for Reactivity Control are potentially inconsistent with the licensee's proposed defueling plan. These TS were not developed with i
defueling as part of their basis, for example, Interim TS 3.1.6.A.1 only allows up to two control rod pairs to be withdrawn for refueling. The licensee's proposed defueling would require more than two control rod pairs i
t withdrawn.
l The licensee has subs.itted by letter dated September 14, 1989 upgraded 15 for Reactivity control.
j 4.0 RE0 VEST FOR ADDITIONAL INFORMATION The staff notes that additional information is requested from the licensee in l
Sections 2.1, 2.2, 2.3, and 2.4 of this preliminary SE. The TS described in i
Section 3.0 have been received from the licensee.
This evaluation is preliminary and will be revised when the requested items are submitted.
Principal Contributor: Kenneth L. Heitner, PDNP Dated: October 17, 1989 l
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