ML20085G333
| ML20085G333 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 06/08/1995 |
| From: | Fisher M PUBLIC SERVICE CO. OF COLORADO |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| P-95057, NUDOCS 9506200166 | |
| Download: ML20085G333 (8) | |
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16805 WCR 19 1/2;.Platteville, Colorado 80651 4
A June 8, 1995 Fort St. Vrain s
l P-95057 j
U.
S. Nuclear Regulatory Commission j
ATTN: Document Control Desk j
Washington, D.C.
20555 Docket No. 50-267 1
. OF THE 10.CFR 50.59 REPORT OF
SUBJECT:
QUARTERLY SUBMITTAL
- CHANGES, TESTS. AMD EXPERIMENTS FOR FORT
'8T.
VRAIN j
DECOMMISSIONING
REFERENCE:
NRC Letter dated November 23, 1992,'Erickson to j
Crawford (G-92244) i Gentlemen:
l This letter transmits the 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 i
description of each change, test and - experiment as well as a summary of the safety evaluation.
This report covers the period of February 16, 1995 through May 15, 1995.
)
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, j
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."
9506200166 950608
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i PDR ADOCK 05000267 W
PDR i
P-95057 June 8, 1995 Page 2 If you have any questions concerning this report, please contact Mr. M. H. Holmes at (303) 620-1701.
Sincerely, vt/
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M. J.
Fisher Decommissioning Program Director MJF/JRJ Attachment cc:
Mr. Michael F. Weber,' Chief Decommissioning and Regulatory Issues Branch Regional Administrator, Region IV Mr. Robert M. Quillin, Director Radiation Control Division Colorado Department of Public Health and Environment
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i June 1995 QUARTERLY 10 CFR 50.59 REPORT OF CHANGES, TESTS AND EXPERIMENTS 4
i FOR FSV D m MMISSIONING J
Background:
j The following is a brief discussion of 10 CFR 50.59 changes to the i
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 February 16, 1995 through May j
15,~1995.
While this report is similar to past reports of changes, tests and 1
experiments submitted in accordance with 10 CFR 50.59, the quarterly decommissioning reports are submitted. pursuant to j
Paragraph (b) (2) of the FSV Decommissioning -Order. (issued' in. NRC i
letter dated November 23,
- 1992, Erickson to Crawford),- which J
states:
I i
i "The licensee shall submit, as specified in 10 CFR.50.4, a I
i report containing a brief description of any. changes, tests j
.and experiments,. including a summary of the safety evaluation 1
3 of each.
The report must be submitted quarterly."
l
.Q.hances to the'FSV Facility or its Procedures as Described in the j
Decommissionina Plan There were no changes to FSV facility procedures, as the procedures i
are described in the DP, during this reporting period.
Changes.to i
plans for decommissioning the facility are described below.
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1.
Installation of Additional Domineralisers in the Drain.Line of the PCRV Shield Water. Polishing Demineralisers-1 i
Section 3.3.2.2 of the DP states: " Water to be' released from the i
i PCRV will be transferred through the PCRV. shield water system demineralizers,' either to a liquid waste holdup tank in the -
existing Radioactive Liquid - Waste System (System 62) or to the 4
for sampling and analysis...
If j
further processing of the radioactive liquid is desired prior to release, water from a system 62 holdup tank can be circulated through the System 62 domineralizers and returned to~the holdup tank and water in the RBS can be circulated through the PCRV-shield f
i water system demineralizers and returned to the RBS."
i This activity installs additional demineralizers in the. drain line of the existing PCRV shield water polishing demineralizers to increase the flexibility and capacity for processing water that I
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originated in the PCRV.
This enables water from the RBS or RBS keyway liner (referred to as the bladder) to be recirculated through the additional demineralizers for cleanup, by means of a RBS pump or an additional pump, as an alternative to the existing recirculation path through the shield water system polishing demineralizers.
The additional demineralizers also enable water drained from the PCRV to be further processed prior to deposit in the RBS/ bladder.
With installation of the tie-in to System 62 at the discharge of the Liquid Wast.a Sump Pumps, described in the following Item 2, processing of irater from System 62 may also be accomplished using the additional demineralizers.
The flow path for recirculating water from the PCRV, through the shield water system filters and demineralizers, back to the PCRV, is independent of the flow paths used for recirculating water through the additional domineralizers (from either the RBS, bladder, or System 62), and this activity permits flexibility to recirculate PCRV water through the shield water system while simultaneously processing water from the RBS/ bladder through the additional demineralizers for cleanup prior to discharge.
Although the additional demineralizers permit selection of a number of different processing flow paths for water drained from the PCRV, the radioactive liquid effluent release path described in DP Section 2.2.3.10 remains the s ame.
All radioactive water released from the Reactor Building is discharged via the liquid waste system discharge line.
The probability of a breach of the shield water system is not increased since the additional demineralizers and associated valves to be installed have a nominal pressure rating greater than the system pressure during PCRV draindown via the shield water system, or during recirculation using a RBS pump or additional pump.
Should a breach occur, any release of water would be substantially bounded by the consequences of Loss of PCRV Shield Water Accident previously evaluated in DP Section 3.4.7.
This accident analysis assumed that 423,500 gallons of water are released to the Reactor Svilding with a tritium concentration of 62.4 C1/cc, whereas a much smaller water inventory exists in the PCRV, with a tritium concentration below 4 E-3 C1/cc.
No new accidents / malfunctions are created since failure of the additional demineralizers or associated valves and hosing could only result in spillage of water into the Reactor Building, which would drain into the RBS.
This would not result in an uncontrolled release to the environment.
Likewise, draining PCRV water, or recirculating the contents of the RBS/ bladder using an RBS pump or an additional pump via the additional demineralizers does not introduce the possibility of uncontrolled release to the environment.
Decommissioning Technical Specification 5.4.4 (a),
" Radioactive Effluent Controls Program" requires FSV to have a
program 2
h to assure doses to members of t ereasonably
'e are as low as conforming with 10 CFR 50.36a Offsite Dose effluents radioactive contained in the to be the primary from of public This
- program, continues and no margin Manual (ODCM),
Specification is achievable.
control over effluent releases,any Technical Calculation administrative in the basis of safety defined installation of reduced by this activity.
the t
was concluded thatot constitute an unreviewed safe y it
- above, Based on theadditional domineralizers does n question.
Path From Radioactive Liquid Installation of Alternate Flow from the 2
flow path additional waste sump alternate 62; sump to the an modification provides installed in the drain lice fromWater c liquid waste (System This radioactive domineralizers discussed above,the shield water system p demireralizers. sump is normally pumped to t it is processed for discharge using liquid wastedescribed in DP Section 2.2.3.10.demineraliz in the radioactive System 62 receiver tanks, wherethe System 62 deminera 62 the System Providing f
k s considerable time.
of limited capacity liquid waste sump to the additional processing water in this manner ta e Due to the drain will reduce an alternate flow path from the shield water system s capable of i
With the additional demineral zer tive liquid waste sump, the Syste the in demineralizers processing time.
processing water from the radioac k
62 demineralizers, receiver tan snecessary, and can b d
ided by installation of a tee in the dioactive liquid waste sump pumps.the System The alternate flow path is prov2 inch discharge line fr or the System 62 demineralizers additional New isolation additional the through id waste sump to either the can either be returned to the liquin prepara flowing After tanks.
receiver demineralizers, watersump or routed to the RBS/ bind er acc d
The probability of occurrence of ansed since the new tee, the new tee to the evaluated in the DP is not increa nominal pressure rating in excess connects and hose (hose sump liquid waste piping radioactivetive liquid waste sump additional demineralizers) have a
- valves, of the in the discharge pressureThe maximum capacity of the rad oac i
concentrations of the the radioactive liquid waste sum tritium Additionally, pumps.
gallons.
water from the PCRV processed inare well below th is 1,000 leak in the of a in a maximum volume of 1,000 consequences the Therefore, accident.
alternate flow path, resultingof water leaking into the R water 3
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s previously evaluated in Section 3.4.7 of the DP, " Loss of PCRV Shielding Water Accident."
Potential mishaps associated with. the alternate flow path introduced by this activity would result in a leak of radioactive liquid inside the Reactor Building, similar to the Loss of PCRV Shielding Water Accident, and would not create a new type of i
accident / malfunction not previously evaluated.
All radioactive liquid effluent releases will continue to use the same affluent-I release' path and be procedurally controlled to prohibit the accidental discharge of radioactive liquids to.the environment.
The Decommissioning Technical Specifications requirements for l
radioactive liquid effluent releases are contained in the ODCM.
The margins established in the 0]cM to assure regulatory compliance -
are not impacted. Therefore, no margins of safety in the basis for any technical specification ars reduced by this activity.
Dased ' on the above, it was concluded that installation of the alternate flow path from the radioactive liquid waste sump does not constitute an unreviewed safety question.
j 3.
Removal of the Fuel Storage Wells DP Section 2.3.4.3,
" System 14 - Fuel Storage Facility", states:
"When the fuel storage wells are no longer needed, each of the nine inner storage wells will be decontaminated to the. criteria for release for unrestricted use, surveyed, and the top access plugs
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replaced."
Experience gained in decontamination and survey on the FSV decommissioning project, with equipment such as the Fuel Handling
- Machine, indicates that
.it is not feasible to decontaminate and survey the fuel storage wells (FSWs) in place, 3
and removal and disposal is the. most effective decommissioning j
option.
Each FSW consists of a concentric tank structure, with an inner and outer liner suspended from the refueling floor within a concrete vault.
Three ESWs are located in each of three vaults.
A cylindrical outer steel liner surrounds four inner steel tubes (cloverleaf shape) that were designed to contain the irradiated fuel.
Current planning calls for removal of the nine outer steel liners, with the integral inner steel configuration.
Each FSW L
liner is approximately 47.5 feet long, 54 inches diameter, and weighs approximately 55,000 pounds.
The FSW liners and shield plegs will be removed by the Reactor Building crane.
These additjanal lifts are considered to have a negligible effect on the probability of a drop accident.
All nine FSWs are projected to have less dispersible activity than assumed I
for a single steam generator primary module.
_The consequencer of a postulated drop of a FSW liner would be bounded by those l
previously evaluated in DP Section 3.4.10,
" Dropping of a Steam 4
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i Generator Primary. Module. "
This activity does not create the possibility of a new or different accident or malfunction, since a j
postulated FSW drop accident is similar to the steam generator j
primary module drop accident previously evaluated.
The operations involved in this activity do not impact the basis for any technical j
specification, and no margins of safety are reduced.
i Based on the above, it was concluded that removal of the fuel l
storage well. liners and shield plugs does not constitute an s
4 unreviewed safety question.
1 4.
steam Generator Primary Modules Removal i
An earlier safety evaluation was amended to assess current steam j
generator primary module removal plans that differ from those
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described in the DP.
The earlier safety evaluation, summarized in the quarterly 10 CFR 50.59 report submitted to the NRC in Reference 1,
assessed packaging the steam generator primary modules in j
cylindrical steel containers on the PCRV ledge or on the refueling i
j floor.
This earlier safety evaluation was amended to assess the most recent change in steam generator primary module removal plans, 4
involving removal of the clamp that attaches each of the twelve i
modules to the lower plenum floor, without the modules supported by l
the Reactor Building crane.
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DP Section 2.3.3.11.1 states:
"The steam generator primary 3
assemblies will be rigged (to the Reactor Building crane using i
standard rigging techniques and devices) to secure them before the j
final unclamping or severance cut."
This final unclamping or cut l
involves detaching the lower plenum floor
- clamp, located j
approximately 2 1/2 feet above the bottom of the steam generator l
primary modules.
The primary purpose of the steam generator lower j
plenum floor clamp was to connect the modules to the lower plenum j
floor, providing a seal between the helium circulator inlet and l
outlet plenums, and it was not intended to be a structural support j
for the steam generator primary modules.
There are schedular j
advantages to having the Reactor Building crane.available while l
steam generator primary modules are detached from the lower plenum floor.
Analysis determined that the detached modules cannot j
topple, even without the support of the Reactor Building crane, as discussed below.
f DP Section 3.4.10 evaluates postulated drop of a steam generator primary module in the Reactor Building truck bay.
The probability of occurrence of this accident is not increased since detaching the modules from the lower plenum floor does not involve physical l
handling, and calculations demonstrate the modules are adequately j
supported without reliance on the clamp at the lower floor.
The consequences of a module drop accident are not affected since the i
radioactivity available for release is unchanged by this activity.
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A calculation was performed which indicates that a horizontal force of 2.6 tons would need to be applied to the top of a steam generator primary module to overcome inertial forces and begin to 4
topple a module.
Wh3n this force is removed, a module would right itself, unless the top of a module is displaced 4.5 ft.,
beyond which the module could continue to topple.
This calculation.is conservative in that it does not take credit for the 36 primary closure studs (1 3/4" dia. and approximately 9 inches long) at the bottom of each primary module or the PCRV lower floor.
While the nuts have been removed from the bottoms.of the primary closure 4
studs, enabling each module to be lifted straight out, these studs would resist toppling,-and studs would have to bend'and/or shear before a primary module could topple.
The PCRV lower floor, consisting of a 1" thick horizontal steel plate supported by beams and located 2.5 ft. above the bottom of the primary module, would also substantially resist toppling.
It is considered that the steam generator primary modules are adequately supported against toppling without reliance on the lower plenum floor clamp assembly.
i since the module being lifted out of the PCRV will be carefully observed throughout'the lifting operation, it is not considered i
credible that the crane operator would collide the module being lifted with another module and then continue to push the top of j
that module a distance of 4.5 f t.
PCRV beltline concrete segments i
'will not be removed until such ' time as all 12 of the steam generator primary modules have been removed from the PCRV.
Based on the substantial resistance to toppling without reliance on the clamp assembly, and the distance the top of a module would need to 3
be moved for toppling to
- occur, it is considered that unclamping/ severing the lower plenum floor clamp at the bottom of steam generator primary rodules while they are not supported by the d
Reactor Building crane does not create the possibility of a primary module toppling in the PCRV.
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l operations involved in this evaluation do not change the bases to any technical specifications, and no margins of safety are reduced.
Based on the above, it was concluded that detaching the steam generator primary modules from the lower floor, with the modules not connected to the Reactor Building crane, does not constitute an unreviewed safety question.
1 Tests or Exoeriments Not Described in the Decommissionina Plan i
No tests or experiments were conducted this reporting period that are not described in the DP.
References 1.
PSC letter, Warenbourg to NRC Document Control Desk, dated September 15, 1994 (P-94077) 1 6
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