ML19347F669
| ML19347F669 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/14/1981 |
| From: | Metropolitan Edison Co |
| To: | |
| Shared Package | |
| ML19347F666 | List: |
| References | |
| NUDOCS 8105220395 | |
| Download: ML19347F669 (17) | |
Text
-
a B. H. Grier
-.2 -
4 H
FACILITY MODIFICATIONS l
buring the reporting period a number of facility modifications and projects were undertaken were that recovery oriented. These changes have been subject to numerous in-house and.NRC review sr.3 ions and NRC approval.has been received prior to implementation in accordance with Tech Spec.' 6.8.2.
A summary of,these recovery oriented modifications along with references to key correspondence are listed below.
Where applicable, the. NRC approval letter is referenced.
j Reactor Building Purge - Memorandum and Order, dated June 12,'1980 Order for Temporary Modification of License, June 12, 1980, B. J. Snyder to R. C.
Arnold.
4 Mini Decay Heat Removal System:
B. J. Snyder letter to R. C. Arnold, Amendment of Order, dated November 14, 1980.
i I
Processed Water Storage Tanks:
'J. T. Collins memo to J. J. Barton September 13, 1979 Comments on Design Criteria.
C. K. Hovey to L. H.
Barrett, 4/1/81 LL2-81-0075, response to NRC/TM1 ;
d 0016 2/27/81. TLL 029,1/24/80, R. F. Wilson to J.
T. Collins. TLL 395-10/09/80, G..K. Hovey to J. T.
Collins Response to NRC/TMI-80-026, 2/1/80, J.. T.
Collins to R. C. Arnold.
l Penetration 401 Modification:
L. H. Barrett letter to G. K. Hovey NRC/TML 009, approval of revised design criteria for l
modification of Reactor Building Penetration 401, Dated February 11, 1981.
l BOP Diesel Generator -
B. J. Snyder letter to R.C. Arnold dated l
August 11, 1980, modification of Order deleting operability requirements for the BOP diesel generator.
1 Nuclear Sampling System -
Operational June 16, 1980 - In Quarterly Report I
June 1980. Conference Notes 6-5-79, TM1 Trailer 105 (J. T. Collins attended).
Interoffice Memorandum
(
transmitting approved design criteria for the
{
sampling system, September 14, 1979 (J. T. Collins signed).
Airlock Contamination Control Facility - J. T. Collins to R. C. Arnold, January 30, 1980, NRC/TML-80-017 - Review of Facility Design Description.
R. F. Wilson to J. T.
Collins, March 4, 1980, TLL-103 - Response to above. Completion of this item was a prerequisite for initial containment entry.
This procedure was signed of f by the NRC in March, 1980.
81052203%
i
B. H. Grier The above modifications include a number of Engineering Change Memos (ECMS).
Other changes made to the facility that were not related to specific recovery projects are listed below by ECM number. The ECM number is a number assigned to each modification as a tracking mechanism.
Change ECM #584 Description - The subject change installs a temporary flange in the Cask Storage Borated Water Pool piping so that during recovery, the Unit 2 BWST water can be transferred to either the Unit I cask pool or EPICOR I.
Af ter recovery, the system will be restored to the original configuration.
Safety Evaluation - The subject modification is intended to be a temporary modification during the recovery mode. The installation of the flange does not reduce the integrity of the system. The modification is installed under B31.7 snbsection class III and Quality Control level 3.
The modification meets seismic catagory I design criteria.
Any failure mode is bounded by previous analysis. Therefore the change modification does not increase the probability of occurrence or consequence of an accident analyzed in Chapter 14 of the FSAR, nor does the change introduce the possibility of a new type of accident other than any previously analyze' in Chapter 14 of the FSAR. The margin of safety as defined by the Technical Specification bases is not reduc ed. Therefore, it is concluded that the subject change does not involve an unreviewed safety question.
1
,...~--. - _,
g B. H. Grier Change ECM #715 Description - This change installs lines from the EPICOR 11 system and the
~
SDS system to the processed water storage tanks.
This feeds the ef fluent from these systems to the processed water storage tanks for storage af ter processing.
4 Safety Evaluation Snemary - The above changes will be handling only pro-cessed water from the EPICOR II and SDS system.
Isotopic content of the water will be tested prior to release to the sterage tanks.
System pining is designed and installed to ASME B31.1 st o._o a rd s.
Since the specific activity of the processed water is very low, it is felt that the system does not reduce margins of safety or increase the probability or consequence of an accident as previously described. Therefore this change does not constitute an unre-viewed safety question.
1 l
l i
l
B. H. Gri r Change ECM #606 Description _ - This change modifies the Fuel Handling Building crane by the addition of inching drives. These drives provide fine i
control of the crane for use with the SDS.
Safety Analysis Summary - The inching drives will be provided power from BOP sources and therefore will have no effect on nuclear safety power supplies. There is no fuel in the Unit 11 fuel pool to require restriction of crane movement. Administrative controls will be added to prevent movement of large loads over critical areas of the SDS and SPC systems.
Since there is fuel in the Unit I fuel pool, the inching drives should be tagged to prevent their use in Unit 1.
With the use of administrative procedures the change does not increase the probability or consequences of an accident previously described nor does it create the possibility of an accident of a different type than previously analyzed in the FSAR. The change does not affect the margin of safety as defined in the Technicel Specification Bases.
It is concluded this change does not constitute an unreviewed safety question.
l
..m v,
e
_m,_
w.g---..__n,_m u.
-ry_,.nm, _ _m.m.-.-,_,,,.m.-,
-~w-
..~, _
.-.r,-
. B. H. Grist Change ECM #815 Description - This change installs a motor operated valve (WG-V-100) downstream of valves BS-V-156A and BS-V-157A on BS-P-1 A drainline. This valve will allow remote draining of BS-P-1A.
Under normal circumstances BS-V-156A and BS-V-157A are closed. In emergency ',ituations, WG-V-100 will be closed and then BS-V-156A and Bs-V-157A will be opened prior to com-mencing pumping. When pumping is halted WG-V-100 will be opened remotely to allow draining of the lines to the BS-P-1 A The valves will then be flushed and placed in vault sump.
normal lineup.
Safety Analysis Summary - Under normal conditions, isolation valves BS-V-156A and BS-V-157A remain in a closed condition.
These valves are only opened during emergency conditions requiring draining of the Reactor Building sump. At this time WG-V-100 is closed and remains closed until pumping operations are complet ed. Valve WG-V-100 is then opened to provide remote draining of the WG-P-1 pump system. This reduces radiation exposures to workers that would otherwise result during manual draining. The primary safety concern for this modification is leakage from valve WG-V-100 when BS-V-156A and BS-V-157A are open. Any leakage from valve WG-V-100 is routed to the BS-P-1A sump via floor drains. Ventilation to the cubicle can be isolated, reducing airborne releases.
Any leakage is bounded by accidents j
previously analyzed in the FSAR.
It is therefore l
felt that this change does not constitute an unreviewed safety question.
t I
l
. B. H. Crist Change ECM #649 Description - This change installs a catch system to provide drainage from the The enclosed BWST enclosed area to the Auxillary Building sump.
areas consist of a concrete base pad, steel frame work and fiber-l glass roofing and siding. Each enclosure has a single floor drain with two inch stainless steel piping connecting the enclosure drains to the four inch BWST drain line. The four inch drain line runs to the Auxiliary Building sump.
Safety Analysis Summary - The enclosures are designed to be Seismic Category I with the base pad at least six inches above ground level. The rest of the enclosure is steel framing, metal doors, and fire resistant fiberglass panels. The building is moisture tight to the extent th at all seams are caulked except a few conduit and piping penetrations which will be sealed. The drain system is not Seismic I since it ties to the BWST drain line below the isolation valve. The fire barrier penetration has been examined to determine if the safety margin of fire protection is decreased. It has been determined that fire protection has not been decreased. The possibility for a previously undescribed accident has not been increased and the margins of safety as defined in the Technical It is Specification Bases has not been reduced.
concluded that this change does not constitute an I
unreviewed safety question.
l l
l l
B. H. Grier Change ECM #704 Description - This change installs a four inch domestic water branch line from the three inch supply header to the search entry facility, administration building, TLD facility and the security I
administration building. The change provides domestic water service to the above buildings.
Safety Analysis Summary - The change deals only with the domestic water supply.
There is no impact on nuclear safety related equipment.
There is no effect on the safety analysis as described on the FSAR and no change to the bases of the Technical Specifications.. It is cot eluded that this change does not constitute an unreview d safety question.
?
l,
-l 1
a e
i l
i g-4 I
i.
I B. H. Grict
' Change ECM #622 Description - This change provides for a slight realEgnment of sone pipe support mounting plates in the Fuel Handling Building. During original installation, the holes drilled for mounting encountered rebar.
This change provides for readjustment of plate position to facili-tate installation.
Safety Analysis Summary - The relocation of the mounting brackets is bounded by the pipe support analysis. There is no ef fect on the safety margins, Technical Specification Bases or accident an alys is.
It is concluded this change does l
net constitute an unreviewed safety question.
L i
P 1
l I
I l
l
Q. H. Crier Change ECM #628 Description - This modification installs a temporary vent in the Nuclear Nitrogen System line between NM-V40 and NM-R9, a relief valve test connec-tion. This change will pressurize the pilot sect ion of NM-R9 whi.h'will prevent an inadvertent lifting and discharge of NM-R9 to the Auxilary' Building environment when the system pressure is raised from 0 psig to operating pressure.
Safety Analysis Summary - This modification will prevent an inadvertent li f t ing of Nit-R9 during system pressurization.
Since the system is radioactively contaminated and the discharge is to the Auxilary Building environment, preventing this release is in the int.erest of the public heslth and safety. - The modification does not prevent th e system from functioning normally during periods of overpressure. Since the system still functions as intended, this modification does not change any safety analysis or bases.
Therefore, this change does not constitute an unreviewed safety question; w
.s
+
4 S
B. H. Grict Change ECM #551 Description - This change removed HP-R-215 from the spent fuel bridge when the bridge was disassembled for installation of the SDS system.
HP-R-215 was an area radiation monitor used as a criticality alarm.
Safety Analysis Summary - The removal of the spent fuel bridge and the associated area radiation monitor HP-R-215 was necessary for installation of SDS equipment.
HP-R-215 served as a criticality monitor. Since no spent fuel will be stored in this araa during SDS operation and since the SDS system will have its own radiation monitoring system, removal of this instrument will not degrade nuclear safety. Removal of this alara will not increase the probability or consequence of a previously analyzed accident, will not create a new catagory accident or will not affect the bases of the Technical Specification. Therefore it is concluded that this change does not constitute an unreviewed safety question.
i
. B. H. Crier -Change ECM #645
. Description - Thic change modifies the Auxiliary Building 'amd Fuel Handling Building exhaust filter housing' and ductworks.
This will allow the installation of injection and sample ports for DOP and
~ Freon Testing.
Safety Analysis Summary - The modifications ensure that proper testing of the filters can be accomplished per Reg. Guide 1.52 and ANSI N510-75. This enhances environmental safety.
Nuclear safety is not adversely affected by this ch ange. The margins of safety discussed in the FSAR or Technical Specification Bases are not affected.
Therefore, this change does not involve an unreviewed safety question.
)
i
...~.n-
4 B. H. Grist
. 4 Change ECM #765 Description - This change installed a jumper between terminals A6 and A7 on all 4160V and 6900V "69" switches. _ The jumper eliminates a local switch that is not alarmed or locked. - This would eliminate the possibility of this switch being in the wrong position and thereby not being able to perform its intended safety function. The permissive function of this switch is also provided by racking in or out the associated breaker.
Safety Analysis Summary - This modification eliminates a generic concern addressed in I & E Bulletin 80-20. The change eliminates the potential for an unalarmed, unlocked switch being in a wrong position, thereby disabling its safety function.
This enhances nuclear safety. It is concluded that this mod _ does not change any present accident analysis, l
modify Technical Specification bases, or introduce a new type of accident. Therefore this modification does not involve an unreviewed safety question.
4 i
l 4
l t
(
1 6
t 4
i
,--r.
w.,-,
.--,_._,*,.-w.
en...,4v..,,--,-,_..,,
-,,,,,.-,.,.,..w--..wm.w.,-.-,
,y.,%y,wo e.yvw,+,y,-y,4
--.--c-,-n,.,-
r.-
B. H. Grier 14 -
Channe ECM #572 Description - The subject modification replaces a gas monitor liner assembly in H2-R-229 with a High Range Noble Gas Monitor assembly. This change provides.the. capability of measuring. noble gas Kr 85 concentrations up to 103 uCi/cc during the Reactor Building Purge..
Safety Analysis Summary - This modification expands the capability of the present monitor in HP-R-229 to detect noble gases.
The change enhances nuclear safety by extending the measurement range of gaseous effluents. The change will not affect the accident analysis as described in the FSAR, nor does it af fect the bases of the Technical Specification. Therefore it is concluded that the change does not involve an unreviewed safety question.
p.
I i
i l
l l
l l
, ~ -, -. - -
,--,,,__----,7g-
,,_y y
e
B. H. Grict Change ECM #817 Description - This change installs a temporary line from the Reactor Building basement area to the Fuel Pool Waste Storage system. This system will be used to transfer water' from the Reactor Building surface suction pump to a storage area to facilitate cleaning the Reactor Building basement.
Safety Evaluation Summary - This change installs a temporary line for pumping out the Reactor Building basement.
It is classified as a liquid radwaste system and as such Reg. Guide 1.143 is imposed. The failure of the piping has been analyzed. It was determined that any breaks are enveloped by other postulated events as described in the FSAR.
The change ooes not affect the bases for an; Technical Specification and as such the margin of safety is not reduced. Therefore, it is concluded that the subject change does not involve an unreviewed safety question.
e B. H. Crist' PROCEDURE CHANGES _
With the issuance of the Interim Recovery Technical Specifications, many pro-i i
cedures issued for surveillance under the Operating Technical Specif cat ons became unnecessary. A large percentage of these procedures were not c Other procedures were unnecessary due being performed due to inaccessability.These procedures were then cancelled and where to the current operational mode.
idance of
~
necessary alternate surveillance procedures were issued under the guAll procedure Recovery Technical Specification Section 6.8.2.
f i
to cancelled received PORC review to determine any impact on sa cancellation.
review prior to cancellation.
a number of procedure changes Additionally, in support of the recovery effort,As required by the Recovery Technical were made and new procedures issued.
NRC Specifications, Section 6.8.2, these recovery related procedures receive Since these procedures have review and approval prior to implementation.
i t
received NRC approval, they will not be discussed further in th s repor.
i l
s 1
I l
f 1
i l
f
j B. H. Crier During the reporting period procedure changes were made to reflect changes to the Unit 2 Recovery organization, communications, facilities and practices.
These changes did not materially change the overall emergency response. Therefore it is felt these changes as listed below, do not constitute an unreviewed safety question.
Approval Procedure No.
Title PCR Number Date 1670.1 Local Emergency 2-80-448 10/10/80 1670.2 Site Emergency Procedure 2-80-449 10/10/80 1670.3 General Emergency Procedure 2-80-450 10/10/80 1670.5 On-Site Radiological Monitoring 2-80-456 10/10/80 1670.6 Off-Site Radiological Monitoring 2-80-457 10/10/80 1670.12 Emergency. Readiness Check List 2-80-467 10/13/80 1670.9 Emergency Training and Emergency 2-80-460 10/10/80 Exercisc Also changed was procedure 1670.4 as described below.
Change PCR 461-1670.4, Radiological Dose Calculation - 10/16/81.
Description of Change - The procedure was updated to reflect ef fluent source terms which Unit 2 would have in an emergency.
It also was updated with respect to ef fluent monitors in use during the recovery period.
Safety Evaluation Summary - This change does not govern equipment operations and is administrative in nature.
It has no impact on nuclear safety and is therefore not considered an unreviewed safety question.
1 i
l l
18 -
B. C. Criss l
TESTS AND EXPERIMENTS During the past year a number of activities revolving around containment entries These activities included sampling of Reactor Building environment,.
occurred.-
decontenination tests, tests on resin column performance, and other tests and for the Reactor Building. These activi--
progr'.ans related to the cleanup effort ties were accomplished using procedures reviewed and approred under the guidance These activities have been also of Recovery Technical Specification 6.8.2.
Since these activities have described in the Recovery Quarterly Reports.
received NRC review and approval prior to initiation they will not be discussed further in this report.
i l
i i
I I
5
-=
~.e-r--
--grr-epm-e
-rv--v6>
- r,t w-+
---,n--,,ci-
,-t+
<m--
-v.,-
-ry,-%
.--m---+v
,wy-e---a