ML19341C222
| ML19341C222 | |
| Person / Time | |
|---|---|
| Site: | Crane, Brunswick  |
| Issue date: | 02/23/1981 |
| From: | Hovey G Jersey Central Power & Light Co, Metropolitan Edison Co, Pennsylvania Electric Co |
| To: | |
| Shared Package | |
| ML19341C220 | List: |
| References | |
| NUDOCS 8103020413 | |
| Download: ML19341C222 (13) | |
Text
o METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LICHT COMPANY AND PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION UNIT II Operat ing License No. DPR-73 Docket No. 50-320 Technical Specification Change Request No. 26 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A/B to Operating License No. DPR-73 for Three Mile Island Nuclear Station Unit 2.
As a part of this request, proposed replaceraent pages for Appendix A/B are also included.
METROPOLITAN EDISON (X)MPANY By
/
W Vice President and Director, TMI-2 I
s e Sworn and subscribed to me this S I day of f r /jnn,ru 1981 G
I$'/$ill9$,&i Notary Public /
J CATHY L. EREY, fictvy PutJic Lcndondttry Tys?.. Dauph;n County. P.s.
Mr Commrss on Expres Oct. 24, iM3 8103020 N
r_
1.
(
Three Mile Island Nuclear Station, Unit 2 (TMI-2)
Operating License No. DPR-73 Docket No. 50-320 Technical Specification Change Request No. 26
' The Licensee requests that the attached changed pages 3.3-8, 3.3-9, 3.3-11, 3.6-1, 3.6-2, 3.7-4, 3.7-5, 3.7-7, 3.9-1, B3 /4 6-2, 5-1, 6-3 be substituted for the existing pages in the Technical Specifications and change pages 2-7, 2-8,. 2-9, 2-10d, 2-11, 2-12, 2-13, 2-14, 2-14b, 2-14c be subst itut ed for the existing pages in the Environmental Technical Specification.
Due to the numerous proposed changes contained in this request each change has been assigned its own subparagraph number for the purpose of maintaining consistency throughout this submittal.
Subparagraph Subject Af fected P ges No.
Appendix A 1
Reactor Coolant Flow Inst rueent ation - Delete 3.3-8 2
Core Flood Tank Level and Reactor Building 3.3-9 Spray Pump Flow Instrumentation - Delete 3
BOP Diesel Fire Detection - Delete 3.3-11 4
Containment Integrity - Correct grammatical 3.6-1 error.
5 Containment Airlocks - Change to Testing 3.6-1 Requirement.
6 Containment Internal Pressure - Increase 3.6-2 maximum;1imit.
~7 Hydrogen Purge Cleanup' System - Delete 3.6-2, B 3/4 6-2 8
. Control: Room Emergency Air Cleanup System - Add a requirement for the control room air inlet radiation monitor to be operable.
3.7-4 9
_ Fire Supression Water System - Change number of High Pressure Pumps required to be operable.
3.7-5
'10 Deluge / Sprinkler System - Permit isolation of Deluge System to filters where charcoal has been removed.
3.7-7 11-
' Fuel Handling' Building / Auxiliary Building 3.9-1
. Air Claanup Systems - Clarify Action Statement 12 Containment Design Pressure and Temperature 5-1 Add occupational exposure considerations.
y 13.
Plant Operations and Review Committee (PORC)-
6-3 change qualification requirements for PORC members.
Appendix B 14 Gaseous Ef fluents - Delete requirement to monitor for iodine 2-7, 2-8, 2-9 2-11, 2-12, 2-13 2-14, 2-14b, 2-14c 15.
Radioactive Gaseous Waste Sampling and Analysis Program. Sampling Location correct ion.
2-10d
-16.
Radioactive Gaseous Waste Sampling and Analysis Program - Delete LLD specification for gaseous emissions.
2-10d Reason for Change Based on an indepth review of and the experience gained in
- nplement ing the u
Recovery Technical Specifications, we have ident ified several changes that should be made to the Technical Specifications to reflect current plant conditions. The reason for each change is identified below with a corres-ponding safetyJevaluation given in a separate section of the report, the specific change is shown on the revised Technical Specification pages which are appended to the end of this request.
- 1) The Reactor Coolant (RC) Flow instrumentation listed on Table 3.3-9 of the Technical Specifications, measures' forced circulation of the reactor coolant. Forced circulation via the reactor. coolant pumps is not a method that will be utilized for maintaining cold shutdown. Therefo re, it serves no functional purpose and should be deleted as other instrumentation is available which adequately monitors reactor coolant conditions.
- 2) Table 3.5-10 should be revised to eliminate the need for Core Flood Tank Level and ' Reactor Building Spray Pump Flow Instrumentation. The systems monitored by these Instruments have previously been deleted from the Tech-nical Specifications and therefore the instrumentation monitoring these systems should also be deleted.
- 3) Table 3.3-11 should be revised to eliminate the BOP Diesel fire detection instrumentation as there is no longer _ any safety-related equipment in the areas they monitor. This item was overlooked in Technical. Specification
. Change. Request Number 22 regarding the' BOP Diesels which was approved by. the NRC as indicated in the Modification of Order dated Aug.ust 11, 1980.
- 4) ' Sectio'n 3.6.l.1 contains a grammatical error in the ACTION statement as it joins an."either" clause with an "and" and could cause confusion. The attached fchange ' corrects this error and eliminates any confusion.
il KZ
e
- 5) Section 3.6.1.1 regarding Containment Air 1.ocks requires an overall air-r lock leakage rate of less than or equal to 0.05 La at Pa, 56.2 psig. This leakage rate cannot presently be verified at TM1-2 because much of the I
preparation of the air lock for this test must be performed from inside the reactor containment building and would result in unacceptably high radia-tion exposures to personnel.
Specifically, in order to perform this test 14 clamps located on the containment side of the airlock door must be installed.
Under normal shutdown conditions this task requires 2 men 2-3 hours to pe rfo rm.
The current condition inside containment will significantly j
increase the time required to perform this task because all personnel entering containment must wear protective clothing and air purifiers thus i
reducing their ef fectiveness. Additionally if for any reason the other l
i-containment air lock became inaccessible or inoperble while the airlock to
[
be tested was clamped shut then any personnel inside containment would be subject to a significant exposure while the airlock to be tested was un-j clamped and opened. The testing requirements in the Technical Specification have been revised to allow testing of airlock components which can be verified without subjecting personnel to the high radiation levels in side L
containmer.t and still provide adequate assurance of airlock operability.
}.
6)
Section 3.6.1.4 timits maximum primary containment pressure to < 0 psig.
This upper limit can now be relaxed to + 1 psig because the reactor building purge has removed the airborne concamination inside containment to the extent that normal leakage from containment will not affect the health and safety of the public.
4 7)
Section 3.6.4.3 Hydrogen Purge Cleanup System can be deleted from the Technical Specifications. This system was needed for the reactor building purge but is no longer used for routine purges of the reactor building
[
which are presently conducted in accordance with NRC approved procedure 2104-4.9* using the reactor building purge system. Permission to perform 2
routine purges.using this procedure was granted by NRC Letter NRC/TM1-80-119 l
from J. T. Ccilins to R.' C. Arnold.
r i
8)
This purpose 'of this change is to ensure that the control room will remain habitable for operations personnel during and following all credible l
accident conditions. Failure of the control room air inlet radiat ion
. monitor could prevent the automatic shif ting of the control room emergency l
air cleanup system to the recirculation mode in the event of an accident.
- 9).
Section 3.7.10.1 should be changed to require only 2 of 4 high pressure pumps to be operable.. This change allows the Unit 1 Circulating Water Flume Diesel. Fire Pump to be removed from service for lack of a water supply with a reasonably assumed requirement to remove an additional fire pump from service for maintenance. These conditions do not impair the Fire Suppression System capability to meet fire flow demand as two pumps
(
supply mora than enough capacity to meet this demand a fact which is already recognized in the bases of Technical Specifications Section 3/4.7.10.
- Therefore the plant is still protected in the event. of a fire.
- 10) Section 3.7.10.2 should be changed to add a # symbol next to *he Reactor Building Purge Exhaust Filters AH-F-31A/B and the Hydrogen Purge Exhaust Filter AH-F-34 to indicate that the fire suppression supply lines to these filters can be isolated by a single manually operated valve. This isolation can be performed because the charcoal has been removed from these filters.
This change does not affect fire detection capabilities in this area, and only changes deluge system operation from automatic initiation to manual initiation. This change also protects against flooding f rom spurious activation of these deluge systems.
- 11) The wording of the ACTION required by Section 3.9.12 for the Fuel Handling Building / Auxiliary Building Air Cleanup Systems shold be reworded to eliminate its ambiguity. This change makes it clear that loss of the Air Cleanup System in one building will not affect movement of radioactive waste in the other building. The recommended rewording is shown on the attached pages.
- 12) Section 5.2.2 cf the Technical Specifications should be changed as indicated The present radiological conditions inside the reactor containment building restricts activity and stay times inside containment. Therefore maintenance inside the reactor containment building must be closely controlled so that only necessary maintenance and retesting is perf ormed in order to effectively control personnal occupational exposure.
- 13) The purpose of this change is to allow the PORC to be compesad of a broader spectrum of qualified personnel then specified in Section 4.4 of AJSI N18.1-1971.
- 14) The Environmental Technical Specifications make repeated reference to I-131 and release rate limits which are not applicable to TMI-2 in its present condition. 1-131 is produced as a result of nuclear fission during reactor operation and has an 8.05 day half life. As the reactor has been shut down since the March 28, 1979, incident. I-131 generation has ceased and I-131 concentration present at that time has decayed to 1 css than detectable levels; therefore, I-131 no longer needs to be monitored. Additionally, weekly monitoring of the charcoal in the Unit Exhaust for I-133 and I-135 can be discontinued for the same reasons as I-131 because these isotopes have shorter half-lives than I-131.
As a result of the radioiodines decaying to less than detectable levels and based on the current and projected inventory of radionuclides at TMI-2, exposure to the thyroid is no longer a critical dose pathway and whole body exposure is now more limiting. Therefore, the proposed Technical Specification change eliminates specific reference to monitoring I-131 and shif ts the emphasis of concern from thyroid exposure to whole body exposure. Particulates with half-lives greater than eight (8) days will still be monitored however.
- 15) Table 2.1-1 requires a monthly grab sample of the EPICOR-II ventilation exhaust. Note b of Table 2.1-1 which clarif f es this requirement specifies that this sample is to be taken f rom the ventilation exhaust from the spent fuel pool area. This note should be corrected as indicated te specify that the grab sample is to be taken from the exhaust of the Chemical Cleaning Building as this is the ventilation system associated with the EPICOR II system.
- 16) Table 2.1-1 specifies that an Analysis for Principal Gamma Emitters be performed on the EPICOR II Ventilation for gaseous releases, and requires analyses for specific gaseous and particulate emissions. The radio-nuclides which are specifically required to be monitored (Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138, for gaseous emissions) are all rela-tively short-lived; 5.27 days for Xe-133 is the longest half-life for these radionclides, these isotopes have decayed to less than detectable levels.
Therefore, the specific requirement to monitor for these radionuclides can be deleted from the Technical Specifications.
Safety Evaluation Justifying Change
- 1) Reactor coolant flow is measured by a dif ferent ial pressure transmitter which measures the pressure drop across a Gentile tube located in the primary piping. This instrumentation is not a reliable indication of the coolant flow through the system while in the natural circulation mode because of the lov flow rates involved and presently serves no purpose for the operator. Additionally none of the proposed modes for maintaining cold shutdown includes forced circulation of the reactor coolant with the reactor coolant pumps and the requirement to maintain the reactor coolant pumps in an OPERABLE standby status was deleted from the Technical Specification by an Admendment of Order dated November 14, 1980. Therefo re removal of this instrument from the Technical Specifications does not affect safety as the remaining remote shutdown instrumentat ion gives suf ficient indication that the reactor remains in a shutdown condition, and adequate instrumentation exists to monitor reactor coolant conditions.
- 2) The Core Flood Tank Level Instruments measure the level in the Core Flood Tanks. These tanks are not required by the Technical Specification to be operable for injection of borated cooling water into the primary system because other means are available which are more than adequate. Therefore these level gauges are not needed for plant safety and should be removed from the Technical Specifications.
The requirement to maintain the Reactor Building Spray System was deleted from the Technical Specifications when the Recovery Technical Specifications were issued. The requirement to maintain the Reactor Building Spray Pump Flow Instrumentation was apparently overlooked when this change was made.
Therefore this instrument should be deleted from the post-accident monitoring instrument list to correct the oversight.
, 3) The basis for listing some fire detectors in the Technical Specification is to ensure prompt detection of fires in areas containing safety related equipment in order to reduce the potentia' i;r damaging this equipment.
As no safety-related equipment is locar-3
.t the area protected by the
-BOP diesel fire detectors listing t!-rar der-ctors in the Technical Specifications is not required.
- 4) This proposed cnange does not affec5 safety is it is only a grammatical correction.
v
- 5) The containme:.t eirlocks are installed to allow access to the reactor building while maintaining a barrier against the possible release of high levels of airborne contamination to the environment.
Prior to purging the reactor building high levels of airborne contamination did exist inside the reactor building and airlock leakage could have resulted in unacceptable releases to the environment.
Since the purge however airborne contaminat ion levels have been greatly reduced and are no longer as great a cause of concern. Therefore the basis for maintaining containment air lock oper-ability is not as critical as when the Recovery Technical Specifications were issued. Additionally the TMI-2 reactor core is in a cold shutdown condition in that Kef f f,0.99 and average coolant t empe rat ur e isf,2000F a mode which normally does not require testing of the containment air locks.
The unique condition of TMI-2, in that there are measurable levels of airborne contaminat ion in the reactor building, approximately 600,000 gallons of highly contaminated water in the reactor building sump, and the core in an unknown configuration, however requires prudent safeguards to be fo llowed. Therefore some testing of the containment airlocks should be pe rfo rmed. Present radiological conditions inside the reactor building greatly restrict access to the building however, thus preparing the contain-ment air lock door for normal type B testing, which requires the installation of 14 clamps on the containment side of the airlock door which normally requires 2 men 2-3 hours to perform, is undesirable from an ALARA standpoint.
As an alternative we will continue to test the airlock door seals as cur-rently required by the surveillance requirements contained in the Recovery Operat ions Plan. This will ensure that the airlock seals are funct ional and free of defects, and due to the controlled process of containment entries any funct ional problems with the airlocks are promptly report ed.
These precautions will enable us to detect airlock degradation so that nec es sa ry repairs can be performed. Addit ionally the airlock serves as a double barrier and any leakage from containment will be mitigated by the airlock and processed through ventilation filters prior to release to the environ-meat.
6)
The maximum containment pressure was limited to f,0 psig af ter the accident due-to very high airborne contamination levels inside the reactor contain-ment building. Since the issuance of the Recovery Technical Specifications the containment has been purged of this airborne activity. Thus allowing cont ainme nt pressure to increase to greater than 0 psig would not af fect the health and safety of the public under existing plant conditions. There-fore we propose to increase the maximum containment pressure limitation to
+1 p,sig.
An analysis performed to determine the of fsite exposure resulting from allowing containment pressure to increase to 1 psig was performed utilizing the following initial conditions.
a)
Containment Leakage Rates =.13 Weight pe rc ent / day This is La at Pa, l
(56.2 psig,) and is therefore a conservative value for leakage at 1 psig.
b)
Containment volume = 2 x 106 ft3
c)
Annual average X/Q at nearest site boundary = 2.3 x 10-6sec/m3
'Ihis value is given in the TMI Of fsite Dose Calcul-Nn Manual and is the value for X/Q used in the TEIS.
d)
Annual Average D/Q at nearest site boundary = 6.5 x 10-8
-2 this is the value for D/Q given in the TMI Of fsite Dose Calculation Manual.
3 uCi/cc for Tritium (Hgurge = 3.5 x 10-activity since the e)
Highest measured contairunent uCi/cc for Kr85 and 8x10-6
), all other isotopes are much less than Kr 85.
It is unlikely that this value will be exceeded under non-accident conditions as current conditions at TMI-2 allow us to purge Krypton from the reactor building on a routine basis and ensures that a significant amount of airborne activity does not build up inside containment.
f)
Total body dose factor for Kr85 = 1.6. x 101 mrem /yr per uCi/m3 -
this value is given in the TMI Of fsite Dose Calculation Manual.
3 for the g)-
Dose parameters for H3 = 6.5x102 mrem /yr per uCi/m 3 2 inhalation pathway and 2.4 x 10 m -mrem /yr per uCi/sec for the food and ground pathways. These valves are given in the TMI of fsite
-Dose Calculation Manual.
The formula's given below were used to determine the of fsite dose from containment leakage at 1 psig and were taken from the TMI Of fsite Dose
. Calculation Manual.
. Dose = Ki (X/Q)v Qiv - for Noble Gases Dose = Pi Dv Qiv - for iodines and particulate isotopes Ki.
= Total body dose factor. due to gamma emissions for each identified noble gas radionuclide.
(X/Q)v = Highest calculated annual average realative concentration for any area e' or beyond the unrestricted area boundary in' 3
sec/m,
r Pi =
Dose. parameter for radionuclides other than noble gases Dv =- - Highest annual average dispersion parameter for estimating the dose to the critical receptor.
. 'Qiv = Release rate of radionuclide i in uCi/sec.
I'
. The. results of the analysis show that the ~ of fsite dose due to raising.
l' the maximum allowed containment press'ure to 1 psig was ~1.2 x.10-4 mrem /
- year which is insignificant in comparision with normal background radiation and will not be a hazard to the health and safety of the public.
~ ~1he possibility exist s for an ay " dent to occur inside containment which would increase the airborne contamination level in the containment above the levels analyzed above.
The' consequences of these accidents are bounded by.
analysis performed in the,TMI-2. FSAR and, the PEIS.and need not be. discussed
~
,here.
t:
Increas ing cont ainment pressure will also increase the pressure on the base of the reactor building over that which already exists due to the anproximately eight (8) feet of water covering the base of the reactor building. This would slightly increase the driving head of any leakage
.from the reactor building into the environment.
Flowrate through the groundwater system is derived from the following formula.
Q=
.A Where Q = Flow Rate K = Permeability i = Hydraulic Gradient A = Cross-Sect ional area of the crack For. the purpose of this equation, the hydraulic gradient is defined as the dif ference between the reported water level in containment and an average low ground water level measured in the monitoring wells adjacent to the containment.
This head dissipates in the rock af ter passing through the containment wall.
This dissipation of the head is caused by the increase in the af fected area as the plume moves away from the postulated crack. Thus increasing containment pressure to 1 psig would not substantially change the period of time required for'this leakage to reach the Susquehanna River. In.this case' the groundwater monitoring wells would provide an early indication of any. leakage so that the consequences of this event can be mit iga t ed. Additionally the NRC staf f analysis referenced in Section 6.3.4.2 of the draf t PEIS indicated that for the act ivity levels found in-
.the sump water that the concentration for Cs-137, Sr-90, and tritium at the nearest' potable. water supply as a result of a leak would be less than 10
~
percent of the MPC for these isotopes. Thus the health and safety of the public would not be endangered.
An additional concern in raising reactor building pressure is that reactor building penetrations 401 and 626 have been altered siace the accident.
Penetration 401 is presently used as an access to the containment for-an apparatus that is.used 'to measure reactor building sump level. This temp-orary arrangement has been satisfactorily tested at 3 inches of mercury.
(approximately 1.5 psi). Additionally we ' intend to modify-this temporary installation to withstand a higher pressure. Therefore allowing ' cont ainment
~
pressure to increase to 1 psig will not interfere with obtaining these measurements or jeopardize the integrity of this installation. Pene trat ion
- 626 is primarily used as 'an access into containment for a radio antenna to
- maintain commucations: with personnel performing reactor building entries.
This installation has been pressureitested to' 56 inches of water (approxi-mately 2 psi).1Therefore allowing' pressure to increase to l' psig will not affect use of.this installation.
7)
The present plant technical specificat ions require the Hydrogen Purge Cleanup System to be operable during purging of the Reactor Building. This system is not used, however', for the routine purging conducted in _accordance
with approved procedure 2104-4.91, therefore the requirement for the Hydrogen Purge Cleanup System to be OPERABLE is no longer applicable aad can be deleted from the technical specifications without af fecting safety.
8)
This requirement increases the safety of control room personnel as it specifically requires that the control room air inlet radiation monitor to be OPERABLE or to place the ventilation system in a recirculation mode.
This ensures that increased radiation levels in the control room air intake will not go undetected or in the event of a failure of the radiation monitor that personnel will not unnecessarily be exposed to higher levels of airborne contamination.
9)
The design basis for the fire suppression system is to transport 2575 gpm to the most remote deluge valve plus 1000 gem to hoses with no more than a 25 psi pressure drop. This volume can be delivered by any 2 of the 4 high pressure fire suppression pumps all of which are rated for 2500 gpm flow rate which is more than adequate to meet design basis fire suppression system demand.
- 10) ' As charcoal is not installed in either the Hydrogen Purge Exhaust Filter
.or the Reactor Building Purga exhaust filters these filters are no longer fire hazards. Therefore isolation of the deluge systems to these filters
- which have had the charcoal ~ removed from them will not affect plant safety.
This will place the deluge system for these filters in a manual mode of operation rather than an automatic mode and prevents a spurious signal from
! needlessly activating the deluge system. Additionally even though the fire supression supply to these filters has been isolated, the fire detectors associated with the deluge systems are not af fected and are still available toLprovide an. alarm in the event a fire should start in the area of the filters.
l') The rewording.of the action required by Section 3.9.12 for the Fuel Handling
. Building / Auxiliary Building Air Cleanup Systems clarifies it s intent and the'refore the possibility of misinterpretation of the intent of this action -
"by the. operator 'is minimized. This change ensures that if either the Fuel Handling Building or.the Auxillary Building' Air Cleanup System is not
- OPERABLE then movement of radwaste in the -affected building is prohibited.
12)[ Due to' th'e present radiological conditions inside the reactor contain-ment building access to and activity inside containment are restricted.
~
l We are aware however of-the need to maintain the containment design speci-ifications itemized in Section 5.2.2 of the' Technical Specifications -and will' perform the' emergency maintenance on the containment th at is necessary
. to achieve this. Additionally we will administratively control those
. repairs that are performed, that under normal conditions should be tested but:because1of existing radiological conditions.inside containment make performance of these ' tests unacceptable from an ALARA standpoint, in order to provide. reasonable assurance that containment design specifications
~
J are maintained, i
w_
- 13). This change 1 enhances plant safety as the ' qualification criteria specified in' Sect ion '4.7.2 of ANSI /ANS-3.1-1978 ~ allows for a broader range of disci-plines and. experience than that specified in Section 4.4 of ANSI N18.1-1971.
(
~
m
14) 1-131, I-133, and I-135 have decayed sit. e the March 28, 1979, incident, approximately 80 half-lives in the case of I-131, to the ext ent that these radionuclides are no longer present in me asurable quant it ies.
Therefore, it is no longer necessary to monitor these specific isotopes.
Additionally, based on the current and projected radionuc lide inventory, whole body exposure is now more limiting than thyroid exposure. We will continue to monitor the activity of particulates with half-lives greater than eight (8) days, which will ensure that the health and safety of the pub lic is not j e opard ized.
- 15) No safety evaluation is required as this change only represents a correction to the sampling location specified in Note b of Table 2.1-1 of the Environ-mental Technical Specification.
16)
Kr-87, Kr-88, Xe-! 33, Xe-133m, Xe-135, and Xe-138, have decayed s ince the March 28, 1979, incident to the ext ent th at these radionuclides are no longer present in measurable quantities. Therefore, it is no longer necessary to monitor these specific radionuclides. We will continue to monitor gross gaseous activity, however, which will ensure that the health and safety of thepublic is not jeopardized. And in the event of a signi-ficant increase in gross gaseous activity, analyses for these radionuc lides will be pe r fo rmed.
Admendment Class The licensee has determined that because the admendment reque st involves several unrelated Class III safety concerns it represents a Class IV License Admendment (per 10 CFR 170.22).
Therefore, enclosed please find a check in the amount of $12,300.00.
Recovery Operations Plan Change Request No. 6 The licensee requests that the attached changed pages 4.1-2, 4.3-7, 4.3-8, 4.6-1, 4.6-2, 4.6-3, be substituted for the existing pages in the Recovery Operations Plan.
These changes that.are proposed based on Technical Specification Change Request No. 26 (TSCR-26) are listed below and cross-referenced to the applicable subpara-graph.at TSCR-26.
' Af fected Page(s)
Subject TSCR-26 Subparagraph No.
~4.3-7 Reactor Coolant Flow Instrumentation 1
4.3-8 Core flood Tank Level and Reactor Building Spray Pump Flow Instrumentation 2
4.6-1 Containment air Locks 6
4.6-2,'4.6-3 Hydrogen Purge Cleanup System 8
' Reasons for Change In accordance with.the changes being requested for. the Technical Specifications several changes will. also be. required.to the Recovery Operations Plan. The specific : chang'es proposed are indicated on the attached pages. The reason for each change is given in the applicable subparagraph of Technical Specification Change Request No. 26.
An ' additional change to the' Recovery Operations Plan' other. than those
'resulting' from 'the'. Technical Specification Change Request No.' 26 is required.
~
This ' additional changes ^is detailed below.
'l)
Paragraph 4.1.1.1.j of the Technical Specifications should be changed asDindicated on1 the. attached pages to clarif y what - tanks ;in the standby reactor -coolant ' pressure control system are to be sampled and1to change
~
- the manner in 'which the dissolved gas sample is taken.-
The change requested
. for the dissolved gas : sampling method is to take this sample at the point where water is made up to the' reactor coolant system (RCS) and is most
_representat ve of the water'which is added -to the RCS.
i Safety Evaluation Justify Changei L
' They safetyE evaluations justif ying the changes to 'the Recovery Operations Plan made as ' a result' of Technical Specifications-Change Request No. 26'are included in Lthe applicable f subparagraph tof Technical Specifications Change Request. No. 26 as cros s-referenced :above.
4 t
4 k
The safety evaluation for the change that af fects unty the Recovery Operations i
Plan is detailed below.
1)
The changes proposed for paragraph 4.1.1.1.j of the Technical Spec i-fications will clarify which tanks in the standby reactor coolant pressure control (SPC) system are to be sampled for boron this clarification will reduce the chance of an error being made about which tanks to sample.
Sampling the SPC system for dissolved gas helps insure that the dissolved gas concentration in the RCS is minimized. The proposed change requires a single sampling point for the entire SPC system at the point most repre-l sentative of the normal makeup to the RCS insuring that the water added to the RCS meets the dissolved gas specification.
i l
l I
r I
i e
i I
l l-l
~
m c
~.. _.,. __--
--