ML20058L251
| ML20058L251 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 12/09/1993 |
| From: | Crawford D PUBLIC SERVICE CO. OF COLORADO |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| P-93113, NUDOCS 9312160217 | |
| Download: ML20058L251 (10) | |
Text
.
4 f
0 Public Service' t
- ". Box B40; -1_
P.O 16805 WCR 19 1/2; Platteville, Colorado 80651 i
i 1
i i
December 9, 1993 Fort St. Vrain j
Unit No. 1
]
P-93113 i
U.
S. Nuclear Regulatory Commission ATTN: Document Control Desk i
Washington, D.C.
20555 Docket No. 50-267
SUBJECT:
QUARTERLY SUBMITTAL OF THE 10 CFR 50.59 REPORT OF.
l
- CHANGES, TESTS AND EXPERIMENTS FOR FORT ST.
VRAIN DECOMMISSIONING
REFERENCE:
NRC Letter dated November 23, 1992, Erickson to Crawford (G-92244)
Gentlemen:
j This letter transmits the fourth quarterly 10 CFR 50.59 Report of Changes, Tests, and Experiments affecting Decommissioning of the Fort St. Vrain (FSV) Nuclear Station. The attached report includes a
description of each change,. test and experiment as well as a l
summary of the safety evaluation. This report covers the period of l
August 16 through November 15, 1993.
This report is being submitted pursuant to Condition (b) (2). of the
" Order Approving Decommissioning Plan and Authorizing Decommissioning of Facility", transmitted in the referenced letter, which states the following:
"The licensee shall submit, as specified in 10 CFR 50.4, a report containing a brief description of any changes, tests and experiments, including a summary of the safety evaluation of each.
The report must be submitted quarterly."
1400?5 Bg21syg1gagg7
[
f4 R
I) l
l l
l l
'l l
l i
P-93113 December 9, 1993 Page 2 t
If you have any questions concerning this report, please contact-
]
Mr. M. H. Holmes at (303) 620-1701.
l t
Sincerely, i
khW Ntut.&
l D. W. Warembourg Decommissioning Program Director DWW/JRJ t
Attachment cc:
Mr. John H. Austin, Chief l
Decommissioning and Regulatory Issues Branch i
Regional Administrator, Region IV j
Mr. Robert M. Quillin,. Director' l
Radiation Control Division
-i Colorado Department of Health
.r l
i
DECEMBER 1993 QUARTERLY 10 CFR 50.59 REPORT OF CHANGES, TESTS AND EXPERIMENTS FOR FSV DECOMMISSIONING
Background:
The following is a brief discussion of 10 CFR 50.59 changes to the Fort St. Vrain (FSV) facility or procedures as described in the Decommissioning Plan (DP) and tests and experiments not described in the DP, in the time period from August 16 through November 15, 1993.
While this report is similar to past reports of changes, tests and experiments submitted in accordance with 10 CFR 50.59, the quarterly decommissioning reports are submitted pursuant to Paragraph (b) (2) of the FSV Decommissioning Order (issued in NRC letter dated November 23, 1992, Erickson to Crawford (G-92244)),
which states:
"The licensee shall submit, as specified in 10 CFR 50.4, a report containing a brief description of any changes, tests and experiments, including a summary of the safety evaluation of each.
The report must be submitted quarterly."
Chances to the FSV Facility or its Procedures as Described in the Decommissionina Plan Descriptions of changes to the facility and procedures, as described in the DP, are as follows:
1.
PCRV Graphite Removal and Packaging Section 2.3.3.8 of the DP describes the methodology originally 1
planned for removing activated graphite - blocks from the PCRV, drying the blocks and packaging them in shipping containers.
This method involved removal of graphite blocks from the PCRV one at a time in shield bells, with transfer of the blocks using shield bells to a dewatering station and dryer, as necessary, then into shipping containers.
The heavy load drop accident analysis, documented in Section 3.4.5 of the DP, assumes an LSA package containing the heaviest of the large side reflector blocks (weighing 2030 lbs.) is dropped in the Reactor Building truck bay and 1% of the activity in the block becomes airborne and is released, following filtration, from the Reactor Building.
The Westinghouse Team (WT) has revised the graphite block handling methodology to enable loading multiple graphite blocks into baskets submerged in the PCRV shield water.
A basket would then be removed 1
= _.
from the PCRV in the shield bell, and the shield bell would facilitate transfer of the basket into a shipping cask liner in the i
hot service facility.
By packagina the large side reflector and side spacer blocks in shielded shipping casks instead of LSA j
t containers, there is no need to dump the stainless steel pins from the side spacer blocks, nor to segment the large side reflector blocks, as described in DP Section 2.3.3.8.
In addition, WT will i
allow the graphite blocks to drip dry and package them with a qualified absorbent media to capture potential incidental liquids, if required, in lieu of processing them through a dewatering station and dryer as described in the DP.
WT estimates that these revisions to the graphite block handling methodology will result in a dose savings of 35 person-Rem, which represents a 45% dose reduction for this activity.
l The probability of a load drop accident will be reduced due to handling multiple blocks in shipping casks instead of handling blocks in the smaller LSA containers, since the number of graphite l
block handling operations in the truck bay will be substantially j
reduced to approximately 84 cask lowering operations.
The weights of the large shield bell and shielded shipping casks loaded with graphite blocks will be well within the rated capacity of the Reactor Building crane.
The worst case heavy load drop accident scenario would no longer involve dropping of a single large side reflector block, as r
analyzed in Section 3.4.5 of the DP, but dropping of a shipping i
cask full of large side reflector blocks, which generally have higher activity levels than side spacer blocks or core support blocks / posts.
The bounding decommissioning accident remains the fire accident as described in Section 3.4.6 of the DP.
The safety evaluation for the multiple block removal and packaging methodology includes the results of analyses of postulated drops of casks full of activated graphite blocks, assuming that the casks are breached and 1% of the activity inventory in the graphite becomes airborne upon impact. Whereas the DP assumed a filtration efficiency of 95%
for particulate removal by the HEPA filters in the Reactor Building l
exhaust stack, the analyses of multiple block drop accidents takes credit for 99% HEPA filter efficiency.
This assumption is in accordance with the revised basis of Decommissioning Technical Specification (DTS) SR 3.2.3, which states:
l 4
i "The filter penetration and bypass acceptance limits in the surveillances are applicable based on a HEPA filter efficiency of 99%, as assumed in the decommissioning accident analysis."
Changes to the SR 3.2.3 basis and the SR 3.2.3 implementing procedure are discussed in Items 2 and 3, below.
Graphite samples were taken and analyzed from the large side reflector blocks and core support blocks in 1993 to determine actual activity concentrations.
In addition, radiation surveys 2
l were performed to determine radiation levels at the accessible surfaces of these blocks.
Based on data from these sample analyses and surveys, activation concentration estimates - in the graphite
'l have been revised from the Activation Analysis, Appendix II of the j
DP.
The most recent activation estimates were used in. the j
applicable accident analyses involving multiple graphite blocks, postulated cask drop and fire accidents, described in the following paragraphs. While graphite samples and contact exposure rates were i
not taken from the side spacer blocks, primarily due to their i
location, refined predictions of activation levels were made based on data from the large side reflector blocks, whose outer surface is adjacent to the side spacer blocks.
j The Activation Analysis predicted that Co-60, Fe-55 and H-3 comprised over 99% of the total activity in the graphite blocks.
This continues to be the case with the revised activation estimates for the large side reflector blocks and side spamr blocks, although the ratios of these three nuclides differ utom those j
predicted in the Activation Analysis.
For the core support blocks, Co-60, Fe-55 and H-3 comprise 92% of the activity, while Eu-152 and Eu-154 comprise approximately 8%.
In addition to Co-60, Fe-55 and H-3, the following nuclides were considered in the dose i
calculations for reanalysis of heavy load drop and fire accidents, with revised graphite activation concentrations:
C-14, Co-57, Ni-63, Eu-152, Eu-154 and Eu-155.
Dose consequences were conservatively calculated for the following postulated load drop scenarios in the Reactor Building truck bay:
A)
The six most radioactive graphite large side reflector blocks in the PCRV, packaged in the same Type B shipping cask, were assumed to be involved in a drop accident.
Although the Type B casks that will be used have a maximum graphite payload of 7,774 lbs, the analysis conservatively used the actual weight of graphite r
in the six most radioactive blocks, 8,500 lbs.
Activation levels were conservatively calculated based on the average contact l
exposure rate of these six blocks, 186 R/hr.
B) 15,500 lbs. of graphite large side reflector blocks having an i
average contact exposure rate of 125 R/hr, packaged in a Type A i
shipping cask, were assumed to be involved in a drop accident.
i This is conservative since the maximum graphite payloads of the i
Type A casks that will be used are 12,732 lbs (with a 1 inch thick liner) and 15,169 lbs (with a 1/2 inch thick liner), and the assumed 125 R/hr contact exposure rate would actually require the blocks to be shipped in a Type B cask with a smaller maximum graphite payload.
c) 15,500 lbs of graphite core support blocks having an average contact exposure rate of 10 R/hr, packaged in a Type A shipping cask (maximum graphite payload of 15,169 lbs), were assumed to be 3
involved in a drop accident.
l Each of the above postulated accident scenarios was analyzed using the same assumptions stated for the heavy load drop accident in DP l
Section 3.4.5, with the exception of graphite activity inventory and HEPA filter efficiency, as discussed above.
The scenario with the highest dose consequences was determined to be B) above, which I
resulted in calculated doses to an individual at the Emergency Planning Zone (EPZ) boundary of 4.1 mrem whole body and 161 mrem to the lungs (the maximally exposed organ).
These dose consequences are lower than those presented in Section 3.4.5 of the DP even though substantially more activated graphite is involved in the postulated drop.
This is due to the credit taken for 99% HEPA l
filter efficiency, instead of the 95% efficiency assumed in the DP.
l The worst case fire accident is described in DP Section 3.4.6, and involves a 30 minute duration diesel fuel oil fire postulated to j
envelop 230 side spacer blocks packaged in LSA containers.
This fire accident was reanalyzed using the revised estimate of i
l activation levels in the graphite side spacer blocks. Although the stainless steel pins will not be dumped from the side spacer blocks, activity in these pins (primarily Co-60) is trapped in the steel and would not be available for release in the event of a fire.
Aside from the revised activation inventories in the graphite, none of the assumptions for this accident were changed.
The reanalysis determined dose consequences at the EPZ of 105.6 mrem whole body and 176 mrem to the lungs (the maximally exposed organ).
These consequences are lower than those identified in Section 3.4.6 of the DP (121 mrem whole body and 215 mrem to the lungs).
It is planned to package the side spacer blocks, the large side l
l reflector blocks and the core support blocks and posts in shielded shipping casks.
The Type B shipping casks are rated for a 30 minute fire.
The Type A shipping casks are not required to be thermally qualified, but would nonetheless provide better l
protection of graphite components in the event of a fire than LSA l
containers considered in the DP.
Since the Type A cask has a l
maximum payload of 15,169 lbs, which is less than the 26,910 lbs l
weight of 230 side spacer blocks assumed in the DP accident l
analysis, the DP accident analysis continues to represent the bounding accident condition.
It is considered that the probability of a fire resulting in the release of significant quantities of radioactivity has been reduced, due to the use of shipping casks in place of LSA packages, previously assumed.
The safety evaluation concluded that the revised methodology for multiple graphite block handling does not increase the probability or consequences of accidents or malfunctions previously evaluated in the DP, for the reasons summarized
- above, considering 4
(
l l
l
. ~-
Y conservative estimates of activity concentrations in the graphite blocks based on actual graphite samples and radiation surveys, and taking credit for 99% efficiency of the HEPA filters in the Reactor i
Building ventilation exhaust stack.
No new accidents or malfunctions are created as a result of the revised block handling methodology since the same lifting equipment previously described in the DP will be used, loads are within the design capacity of the overhead crane and no new failure modes are being introduced.
No Technical Specification margins of safety are impacted by the i
revised graphite block handling methodology.
Use of the 99%
l Reactor Building ventilation exhaust HEPA filter efficiency is in accordance with the revised DTS SR 3.2.3 basis and the revised SR 3.2.3 implementing procedure, discussed in the following paragraphs.
i In addition to multiple block handling and packaging, the. safety evaluation reviewed not dumping the boronated stainless steel pins from the side spacer blocks, elimination of the block dewatering and dryer stations, and not sectioning the large side reflector blocks.
It was determined that these changes in decommissioning
+
planning do not constitute an unreviewed safety question.
j 2.
Change to the Basis of Decommissioning Technical Specification (DTS) SR 3.2.3 The analysis of the heavy load drop accident (DP Section 3.4.5),
involving drop of a single large side reflector block in the Reactor Building truck bay, assumed 95%
particulate removal j
efficiency of the HEPA filters in the Reactor Building ventilation exhaust systeu.
In order to demonstrate acceptable accident consequences for analysis of drop of multiple graphite blocks to support the graphite removal / packaging / handling scheme described in Item 1,
above, it was necessary to take credit. for ' the higher j
particulate filtration capability of the HEPA filters.
On May 18, 1993, PSC submitted to the NRC (Reference 1) an amendment request to the DTS which would tighten the acceptance criteria for the j
Reactor Building ventilation exhaust system HEPA filters' in-place penetration and bypass leakage test (SR 3.2.3) from less than 1% to less than 0.05%.
Reference 1 also requested an associated change t
to the basis of SR 3.2.3 from an assumed HEPA filter efficiency of I
95% to 99%, as a result of the upgraded in-place-penetration and bypass leakage test requirements.
By October, 1993, the NRC staff had completed its technical and legal reviews, and found the proposed revisions to SR 3.2.3 to be acceptable.
However, the amendment had not been noticed in the Federal Register for a 30 day public comment period, in accordance with 10 CFR 50.91.
This posed schedular concerns, since WT was planning to begin removing activated graphite components from the PCRV in mid-October, 1993.
For this reason, PSC performed the 10 CFR 50.59 safety evaluation on a change to the basis of SR 3.2.3, 5
l l
j i
i I
which revised the HEPA filter efficiency assumed in accident analysis from 95% to 99%.
The definition of " Bases" in the DTS i
states:
"In accordance with 10 CFR 50.36, the Bases are not considered part of the Decommissioning Technical Specifications."
Therefore, this change did not require a change to the Technical Specifications, and could be evaluated under 10 CFR 50.59.
The safety evaluation notes that recent testing of the HEPA filters determined penetration and bypass leakage of less than 0.05%
(greater than 99.95% particulate removal efficiency).
Paragraph C.S.c of Regulatory Guide 1.52 (Reference 2) permits the use of a 99% particulate removal efficiency in accident dose evaluations, provided that testing demonstrates penetration and bypass leakage of less than 0.05%.
The safety evaluation determined that this change would not increase probabilities or consequences of accidents or malfunctions previcusly evaluated, citing the worst case consequences of i
potential accidents involving multiple graphite block handling discussed in Item 1,
above.
Dose cor. sequences to an individual assumed to be standing 100 meters from the Reactor Building from the worst case postulated multiple block drop accident were calculated to be 4.1 mrem to the whole body and 161 mrem to the lungs.
These consequences are less than the 7.1 arem whole body and 202 mrem lung dose consequences of the single large side reflector block drop accident evaluated in the DP, due to credit for 99% HEPA filter efficiency.
The safety evaluation determined that this change does not create any new accidents or malfunctions not previously evaluated, since the activated graphite block drop accident is the only accident l
evaluated in the DP that relies on filtration.
This activity does l
not introduce any new failure modes.
The safety evaluation concluded that no margins of safety defined in the bases of Technical Specifications are reduced.
Use of 99%
l HEPA filter efficiency in accident analyses, when the leakage test l
acceptance criteria of less than 0.05% is met, is consistent with the recommendations of Position C.S.c of Reg. Guide 1.52 (Reference
- 2) and NUREG-1431 (Revision 0), " Standard Technical Specifications, Westinghouse Plants," September, 1992.
As stated in the above discussions, the basis for SR 3.2.3 was changed since the HEPA filters had been demonstrated by testing to have an efficiency greater than 99.95%, and the acceptance criteria in the SR 3.2.3 i
implementing procedure had been revised to assure the filters will be required to maintain greater than 99.95% efficiency in order to pass future SR 3.2.3 surveillance tests.
Subjecting the HEPA filters to the >99.95% efficiency acceptance criteria, instead of i
>99%, provides assurance that a greater fraction of particulates will be removed prior to offsite release in the event of a load drop accident involving activated graphite blocks, and any other decommissioning accident that occurs when the Reactor Building 6
i l
l
d 4
4 4
i ventilation exhaust system is operating.
This will result in j
mitigation of offsite dose consequences.
]
Based on the above, it was determined that this Decommissioning Technical Specification basis change does not involve an unreviewed 3
safety question.
It should be noted that the NRC issued Amendment 1
No. 86 to the FSV Possession-Only License on November 27, 1993 (Reference 3),
revising the Decommissioning Technical.
Specifications in accordance with the Reference 1 amendment request.
l j
3.
Change to the SR 3.2.3 Implamenting Procedure, SR 3.2.3-1.5YX i
This activity establishes the
>99.95%
particulate removal efficiency acceptance criteria for the SR 3.2.3 HEPA filter leak i
test, which supports credit for 99% HEPA filter efficiency in l
accident dose calculations, permitting revision to the basis of Technical Specificacion SR 3.2.3, as discussed in Item 2, above'.
Procedure SR 3.2.3-1.5YX 1mplements the Reactor Building.
j ventilation exhaust system HEPA filter surveillance test
]
requirements of the DTS.
This surveillance is required to be performed every 18 months, after structural maintenance on the HEPA' i
filter housing, or after each complete or partial replacement of a HEPA filter bank.
Procedure SR 3.2.3-1.5YX was revised to specify an acceptance criteria of >99.95% HEPA filter efficiency (less than 0.05% leakage).
As discussed above, the guidance in. Paragraph
)
C.S.c of Reg. Guide 1.52 (Reference 2), and NUREG-1431 (Revision
?
0),
" Standard Technical Specifications, Westinghouse Plants,"
September,1992, permit taking credit for a 99% particulate removal l
efficiency in accident dose evaluations, provided that testing l
demonstrates penetration and bypass leakage of less than 0. 05%.
l This activity assures that HEPA filters whose particulate removal i
efficiency is not greater than 99.95% will not pass the SR 3.2.3 surveillance test for leakage and will be declared inoperable.
~
i The safety evaluation determined that this activity has no effect l
on the probability of occurrence of accidents or malfunctions, but will assure that the worst case activated graphite drop accident dose consequences are within acceptable limits.
These consequences, discussed in Item 1 above, are less than the dose consequences of the single large side reflector block drop accident evaluated in the DP, due to credit for 99% HEPA filter efficiency.
The consequences are also well below those of the postulated truck fire involving graphite blocks (121 mrem whole body and 215 mrem lungs), which is the bounding decommissioning accident. Specifying an acceptance criteria of >99.95% particulate filtration efficiency of the HEPA filters in SR 3.2.3-1.5YX does not create any new accidents or malfunctions.
The more stringent acceptance criteria does not involve a reduction in the margin of safety of any Technical Specification bases.
Therefore, it was determined that this procedure change does not involve an unreviewed safety question.
7
]
l
.c Tests or Experiments not Described in the Decommissionina Plan i
l No tests or experiments have been conducted during this reporting j
period that are not described in the DP.
5 I
4 I
s i
l i
REFERENCES
]
)
4 l
1.
PSC letter dated May 18, 1993 (P-93046), Crawford to Austin-1 (NRC) ;
Subject:
" Proposed Amendment to Decommissioning Technical'
{
Specifications.
i i
2.
Regulatory Guide 1.52, Revision 2, dated March 1978, " Design, 4
Testing, and Maintenance Criteria for - Post Accident Engineered-i Safety-Feature Atmosphere Cleanup System Air Filtration and i
Adsorption Units of Light-Water-Cooled Nuclear Power Plants."
i 3.
NRC letter dated November 29, 1993 (G-93182), Pittiglio (NRC) l I
to Crawford;
Subject:
" Issuance of Amendment No. 86 to the Fort St.
t Vrain Nuclear Generating Station Decommissioning Technical l
Specifications."
j i
t
{
l l
4 i
1 9
s 8
~. _
. _