ML20205H069
| ML20205H069 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 04/01/1999 |
| From: | Langenbach J GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20205H074 | List: |
| References | |
| RTR-NUREG-1430 1920-98-20681, NUDOCS 9904080046 | |
| Download: ML20205H069 (17) | |
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{ GPU Nuclear,lnc.
( Route 441 South NUCLEAR Post Office Box 480 Middletown. PA 17057-0480 April 01,1999 Tel 717-944-7621 l U. S. Nuclear Regulatory Commission 19?.0-98-2 % 81 Attention: Document Control Desk Washington, DC 20555
Dear Sir:
Subject:
Three Mile Island Nuclear Station, Unit 1 (TMI-1),
Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 262 Sections 1,3, and 4 Improvements In accordance with 10 CFR 50.4 (b) (1), enclosed is Technical Specification Change Request No.
262. Also enclosed is the Certifice.te of Service for this request certifying service to the chief executives of the township and county in which the facility is located, as well as the designated official of the Commonwealth of Pennsylvania, Bureau of Radiation Protection.
The purpose of this TSCR is to request that the TMI-l Technical Specifications Sections 1.4.2,1.4.3, 1.4.4, 3.1.12.3, 3.3.1.2.b and d, 3.3.1.3,3.3.2, 3.3.2.1, Table 4.1-1 (Items 14, 31, and 32), Table 4.1-3 (Item 4),4.5.2.1, 4.5.2.3, and 4.5.3.1, be revised to add LCO action statements and make rurveillance requirements more consistent with the Revised Standard Technical Specifications for B&W Plants (NUREG-1430), to correct conflicts or inconsistencies caused by earlier Technical Specification revisions, and to revise spent fuel pool sampling from monthly and aP.er adding chemicals to weekly.
Pursuant to 10 CFR 50.91 (a) (1), enclosed is our analysis, applying the standards in 10 CFR 50.92 to make a determination of no significant hazards considerations. As stated above, pursuant to 10 CFR 50.91(a), we have provided a copy of this letter, the proposed changes in the Technical Specifications, and our analyses of no significant hazards considerations to the designated representative of the Commonwealth of Pennsylvania.
Sincerely,
/}(h/
7U114 J. W. Langbfibach I
Vice President and Director, TMI VLK
Enclosures:
(1) Technical Specification Change Request No. 262 (2) TMI-l License and Technical Specification revised pages (3) Certificate of Service for Technical Specification Change Request No. 262 cc: Administrator Region 1 TMI Senior Resident Inspector 7
TMI-l Project Manager File 97033 "
9904000046 990401 PDR ADOCK 05000289 j p PDR ,
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UNITED STATES OF AMERlCA NUCLEAR REGULATORY COMMISSION IN THE MATTER OF DOCKET NO. 50-289 GPU NUCLEAR INC. LICENSE NO. DPR-50 CERTIFICATE OF SERVICE This is to certify that a copy of Technical Specification Change Request No. 262 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with executives of Londonderry Township, Dauphin County, Pennsylvania; Dauphin County, Pennsylvania; and the Pennsylvania Departmerit of Environmenta: Resources, Bureau of Radiation Protection, by deposit in the United States mail, addressed as follows:
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Mr. Darryl LeHew, Chairman Ms. Sally S. Klein, Chairman Boaid of Supervisors of Board of County Commissioners L andonderry Township of Dauphin County Dauphin County Courthouse R. D. #1, Geyers Church Road Front & Market Streets Middletown, PA 17057 Harrisburg, PA 17101 Director, Bureau of Radiation Protection PA Delit. of Environm ntal Resources Rachael Carson State Office Building P.O. Box 8469 Harrisburg, PA 17105-8469 Attn: Mr. Stan Maingi GPU NUCLEAR INC.
i BY: um[t ~
Vicd8 resident and Dpor, TMI DATE: ,/!Y7 i
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Enclosure 1 TMI-l Technical Specification Change Request No. 262 l And No Significant Hazards Consideration Analysisl. i 1
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TECHNICAL SPECIFICATION CHANGE REQUEST (TSCR) No. 262 GPU Nuclear requests that the following changed replacement pages be inserted into ,
s the existing Technical Specification:
Technical Specification revised pages: 1-3,3-18e,3-18f,3-21,3-22,4-4,4-6,4-10,4-41,4-42, and 4-43. l
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These pages are attached to this change request.
II. REASON FOR CHANGE WITH BRIEF DESCRIPTION OF CHANGES. SAFETY l EVALUATION JUSTIFYING CHANGE. AND MARKUP One purpose of this Technical Specification change request is to incorporate certain improvements from the Revised Standard Technical Specifications for B&W Plants (NUREG-1430) that would add limiting conditions for operation action statements, make surveillance requirements more consistent with the Revised Standard Tecnnical Specifications, correct !
conflicts or inconsistencies caused by earlier Technical Specification revisions, to revise spent fuel pool samp'ing from monthly and after adding chemicals to weekly, and correct administrative errors.
Another purpose of this Technical Specification change request is to correct an administrative control that was established to reduce the possibility of a low temperature overpressure event but could potentially limit the use of the high-pressure injection to mitigate a design basis accident.
The following refers to page numbers associated with the location of text on the revised pages and describes the changes, provides a safe +y evaluation justifying the change, and includes a markup of the existing specification:
A. Technical Specifications Page 1-3:
. Description The Upgraded Final Safety Analysis Report was revised to eliminate Figure 7.1-1. The recommended changes in Technical Specifications sections 1.4.2,1.4.3, and 1.4.4 replace the references to this non-existent figure and refers to the appropriate UFSAR section.
Safety Evaluation
- These changes are administrative in nature and serve only to correct an omission from a previous Technical Specification amendment. Also the referenced sections of the UFSAR give a more complete description of the Reactcr Protection System including the controlled drawing which constituted the previous figuie 7.1-1.
Markup -
1.4.2 REACTOR PROTECTION SYSTEM The reactor protection system is shown described in Fig = l- 1 Section 7.1 of the Updated FSAR. It is that combination of protection channels and associated circuitry which forms the automatic system that protects the reactor by control rod trip. It includes the four protection channels, their associated instrument channel inputs, manual trip switch, all rod drive control protection trip breakers, and activating relays or coils.
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- l.4.3 PROTECTION CHANNEL A PROTECTION CHANNEL as shewn described in Fi;;urc 7 ! ! Section 7.1 of the updated FSAR (one of three or one of four independent channels, complete with sensors, sensor power supply units, amplifiers, and bistable modules provided for every reactor protection safety parameter) is a combination of instrument channels forming a single digital output to the protection system's coincidence logic. It includes a shutdown bypass circuit, a protection chann, i bypass circuit and a reactor trip module.
1.4.4 !IACTOR PROTECTION SYSTEM LOGIC This system utilizes reactor trip module relays (coils and contacts) in all four of the protection channels as shewn described in Figure ! ! Section 7.1 of the updated FSAR, to provide reactor trip signals for de-energizing the six control rod drive trip breakers. The control rod drive trip breakers are arranged to provide a one-out-of-two-times-two logic. Each element of the one-out-of-two-times-two logic is controlled by a separate set of two-out-of-four logic contacts from the four reactor protection channels.
B. Technical Specifications Page 3-18e and 3-18f:
Descriptien The present Tech Spec 3.1.12.3 prohibits operation of the High-Pressure injection system whenever "the reactor vessel head is installed and T.v is 1332 F " This prohibition is part of the low-temperature over-pressure protection for the reactor vessel. In the case of a loss-of-coolant accident, high-pressure injection i; ssigned to ptovide core cooling until the reactor coolant system pressure is reduced sufficiently for low-pressure injection During a small-break loss-of-coolant accident, reactor coolant system pressure may be greater than that at which sufficient low-pressure injection flow can be maintained and T.v, is 1332 F. In such a case, core cooling should take priority over the low temperature over-pressure protection concern, The recommended change recognizes this potential conflict and adds a condition to the limiting condition for operation that will allow continued high-pressure injection operation, without violating the Tech Specs if high-pressure injec. ion is operating during an emergency cool down.
Safety Evaluation The administrative controls established to reduce the possibility of a low temperature overpressure (LTOP) event were never intended to limit the use of HPI to mitigate a design basis accident. This change clarifies the LTOP controls (1) to reduce the possibility that this specification would cause HPI to be secured and (2) to eliminate the burden on the Emergency Director to invoke 10CFR50.54x during an emergency situation. The LTOP analysis is concerned with events during the normal cooldown condition.
- Not disabling HPI when RCS temperature is less than 332 F is not a direct threat to plant
' safety. With HPI operating, plant procedures require the control room staff to monitor the RCS pressure and temperature relationship, and throttle HPI before the pressure approaches the Technical Specification figure 3.1-1 pressure vs. temperature limits. During an emergency cooldown, with the revised Technical Specifications, HPI would operate as long as necessary to establish stable verified LPI flow and then HPl would gradually be secured (MU-V-16A/D
- would be closed). When the possibility of requiring HPI flow is sufficiently remote, the HPI pumps will be shutdown and the auto actuation by ESAS would be disabled. Making this j change in administrative controls in this Technical Specification is consistent with the TMI ;
Safety Evaluation, SE-000221-006 and SE-115201-012, which provides the followmg ,
description of the TMI LTOP mitigation system (underlining added). !
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" Low temperature overpressure events (LTOP) events are potential occurrences as a result of various abnormal plant maneuvers during the normal heatun and cooldown orocesses or a result of a component failure and which could result in an uncontrolled pressurization of the RCS above the coerating P/T limits [SE-000221-006 and SE 115201-012]."
The sets of events that have been analyzed for the TMI overpressure protection system are as follows:
e Erroneous Actuation of the HPI system
. Erroneous opening of the core flood tank discharge valve e Erroneous addition of nitrogen to the pressurizer e RCS makeup control valve fails full open a ' All pressurizer heater erroneously energized e Temporary loss of decay heat removal
. . Thermal expansion of RCS after starting an idle RC pump.
The analysis for the above listed events was submitted to the NRC and the staff accepted these analyses as stated in SER for the TMI Low Temperature Overpressure Mitigation System -TMI License Amendment number 56 and now appropriately described in the TMI FSAR.
1 During an Emergency Cooldown the Operators would still be expected to change the j PORV setpoint to the Low Temperature Setpoint as described in Tech Spec 3.1.12.2.
The reduced PORV setpoint as well as all other operator actions assumed, as part of the LTOP mitigation system to provide LTOP protection would remain available. The only analyzed design event that is not protected by the PORV low-pressure setpoint (485 psig) is an inadvertent HPI system actuation. Since Emergency Injection due to a j breached Reactor Coolant System pressure boundary would not be inadvertent (erroneous) injection, the Administrative limits to rack out HPI as part of the diverse and redundant LTOP protection should not be invoked during such an event. Operators will actively monitor the RCS pressure and temperature to ensure that the Tech Spec limits to prevent an RCS overpressure condition are met as described in the emergency operating procedures. The LTOP limits were implemented to provide additional !
assurance against a potential overpressure condition during the normal heatup &
cooldown sequence. The present TS could inappropriately encumber mitigation of an event in progress to address a potential event. Operator monitoring of HPI operation provides sufficient assurance that HPI will be controlled and a potential overpressure j event is avoided. RSTS do not address the potential conflict with HPI operation during j an emergency cooldown.
The effect'of this Technical Specification change on the possibility of a rupture of the decay heat system due to high-pressure injection operation was evaluated. Without proper design and administrative controls, high-pressure injection has the potential to overpressurize the decay heat system. There are three significant elements used to reduce the possibility of such a failure to a minimal level. (1) The decay heat system operating procedures includes a precaution to verify that the high-pressure injection valves are closed with the breakers open prior to placing the decay heat system into operation. (2) The DH/RCS isolation valves (DH-V-l&2) will automatically close if reactor coolant system pressure increases to 400 psig. (5) The PORV will open automatically if reactor coolant system pressure increases to 485 psig after the valve has -
been placed in its LTOP mitigation control mode at less than 275 F. These design and administrative controls are adequate to ensure that the possibility of decay heat system rupture is acceptably low.
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Markup i Page 3-18e 3.1.12.3 If the reactor vessel head is installed and Tavg is <332 F, High Pressure Injection Pump breaker shall not be racked in unless:
- a. MU-V16A/B/C/D are closed with their breakers open, and MU-V217 is Closed, and Pressurizer level is <.,220 inches. If pre:surizer level is >220 inches, restore level to .<,220 inches within I he nr.
- b. Pr:= ri :r !: :! i: C20 i .ch=. If pr;=urizer !: vel i: >220 inch =, :=ter: !: vel te s20 inch = vdF- ! hour.
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- b. Emergency plant condition erists that requires High Pressure Injection.
Page 3-18f During normal plant heatup and cooldown, Wwith RCS temperatures less than 332 F and the makeup pumps running, the high pressure injection valves are closed and pressurizer level is maintained less than 220 inches to allow time for action to prevent severe overpressurization in the event of any single failure.
C. Techical Specifications Pages 3-21 and 22:
Descriotion The present specifications 3.3.1.2.b,3.3.1.2.d,3.3.1.3.b, & 3.3.1.3.c do not have associated action statements. The revised specification includes action statements consistent with NUREG 1430, B&W Standard Technical Specifications. STS 3.5.1 specifies a maximum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with CFT boron concentration not within limits. He TMI specification 3.3.1.2.b is the equivalent specification to standard technical specification 3.5.1.A. RSTS 3.6.7 specifies a maximum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with an inoperable spray additive iystem. The TL specifications 3.3.1.3.b & c are the equivalent specifications to RSTS 3.6.7 The present specification 3.3.1.2.d conflicts with the action statement 3.3.2. Specification 3.3.1.2.d requires one CFT pressure instrument channel to be operable. The rr.aintenance spxification explicitly states its applicability to CFT pressure instruments but then allows for 72 hom only as long as it will not remove more than one train from service. The same conflict applies to the level instrument channel. Specification 3.3.2 apparently intended to allow 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when specification 3.3.1.2.d was not met, but the present statement does not clearly allow this. The revised wording would make this clear.
Safety Evaluation Rese specifications do not have associated action statements. He revised specification includes action statements consistent with NUREG 1430, B&W Standard Technical Specifications, and provide assurance of proper actions in the event Wese specifications are not j met.
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Here is no STS requirement for CFT pressure or level instrumentat:on operability. He CFT pressure and level instruments do not have an accident mitigation function. This issue was I reviewed and the NRC agreed with this conclusion when accepting the RG 1.97 categorization of these instruments. These instruments are RG 1.97 Category 3.
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Markup 3.3.1.2 Core Floodina System
- n. Two core flooding tanks each containing 940 30 ft' of borated water at 600 25 l psig shall be available. Specification 3.0.1 applies. i
- b. Core flooding tank boron concentration shall not be less than 2,270-ppm boron. l Specification 3.3.2.1 applies.
- c. The electrically operated discharge valves from the core flood tank will be assured open by administrative control and position indication lamps on the engineered safeguards status panel. R. .pective breakers for these valves shall be open and conspicuously markea. Specification 3.0.1 applies.
- d. One core flood tank pressure instrumentation channel and one core flood tank level instrumentation channel per tank shall be operable. Specification 3.3 2.1 applies. I 3.3.1.3 Reactor Buildina Spray System and Reactor Buildina Emergency Coolina System he following components must be OPERABLE:
- a. Two reactor building spray pumps and their associated spray nozzles headers and two reactor building emergency cooling fans and associated cooling units (one in each train). Specification 3.0.1 applies.
- b. The sodium hydroxide (NaOH) tank shall be maintained at 8 fL A6 inches lower than the BWST level as measured by the BWST/NaOH tank differential pressure indicator. He NaOH tank concentration shall be 10.0 5 weight percent (%).
Specification 3.3.2.1 applies.
- c. All manual valves in the discharge lines of the sodium hydroxide tank shall be
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locked open. Specification 3.3.2.1 applies.
I 3.3.1.4 Coolina Water Systems - Specification 3.0.1 applies.
- a. Two nuclear service closed cycle cooling water pumps must be OPERABLE. ) '
- b. Two nuclear service river water pumps must be OPERABLE.
- c. Two decay heat closed cycle cooling water pumps must be OPERABLE.
- d. Two decay heat river water pumps must be OPERABLE.
- e. T wo re:.ctor building emergency cooling river water pumps must be OPERABLE.
3.3.1.5 Engineered Safeguards Valves and Interlocks Associated with the Systems in Specicications 3.3.1.1,3.3.1.2,3.3.1.3,3.3.1.4 are OPERABLE. Specification 3.0.1 i applies.. {
l 3.3.2 Maintenance or testing shall be allowed during reactor operation on any component (s)in the makeup and purification, decay heat, RB emergency cooling water, RB spray, BWST level instrumentation, or cooling water systems which will 4 not remove more than one train of each system from service. Components shall not be removed from service so that the affected system train is inoperable for more than 72 consecutive hours. If the system is not restored to meet the requirements of Specification 3.3.1 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the reactor shall be placed in a HOT SHUTDOWN condition within six hours. ;
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3.3.2.1 If the CFT borom c::ce:tratics is cuiside cflimits, cr the CFT pressure instrumentation, CFT level instrumentation, or RB spray NaOH addition system is inoperabic, restore the system to operable status within ~2 hours.
If the system is not restored to meet the requirements of Specification 3.3.1 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the reactor shall be placed in a HOT SHUTDOWN condition within six hours.
D. Technical Specifications Page 4-4 (Table 4.1-1, Item 14):
Description Tech Spec 4.1-1.14 (item 14 on Table 4.1-1) was originally (i.e. Amendment 47 issued with FSAR) "High Pressure Injection Logic Channel." It revised TS 14.1-1.14 to read "High Reactor Building Logic Channel" by errors in reproduction when other changes were made to this page. There is no Technical Specification change request or Technical Specification Amendment that supports the present wording. The original wording was correct. There are four items on Table 4.1-1 for quarterly testing of ESAS logic channels (items 14,16,18 & 20).
These correspond to the four major sections of the original ESAS logic. This change would correct an editorial error that revises TS 4.1-1.14 to read "High Pressure injection Logic Channel."
Safety Evaluation This administrative change restores the Technical Specifications to the approved condition.
The present version of this specification was not intentionally changed. This change corrects an editorial error made in the past.
Markup
- 1. TABLE 4.1-1 (Continued) 2.
CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS
- 8. High Reactor Coolaat S M R Pressure Channel
- 9. Low Reactor Ccolant S M R Pressure Channel l
- 10. Flux-Reactor Coolant Flow S M F Comparator
- 11. (Deleted) - - -
- 12. Pump Flux Comparator S M R 1
- 13. High Reactor Building S M F Pressure Channel
- 14. High Reacter Bu!! Sag NA Q NA Pressure Injection Logic Channels E. Technical Specifications Page 4-6 (Table 4.1-1 items 6,31, and 37).
Descriotion License Amendment No.196 (dated 9/19/95) revised the T :chnical Specifications to remove the Makeup, Purification, and Chemical Addition syster from Section 3.2 of the Technical Specifications and the pertinent design information was relocated to the UFS AR. This proposed change serves to conform the Technical Specifications to
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Amtndm:nt No.196 due to an administrative ovtrsight in the corresponding TSCR No.
252. Therefore, the change is administrative in nature.
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. This proposed change removes instrument surveillance requirements for the Boric Acid Mix Tank temperature instrument and the Reclaimed Boric Acid Storage Tank temperature instrument and is consistent with the Standard Technical Specifications for Babcock and l Wilcox Plants, NUREG-1430, July 1992. Also this change meets the intent of the Final l Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors issued July 1992 and codified in 10 CFR 50.36.
Safety Evaluation 1
These items were changed to correct an administrative error. License Amendment No.
196 (dated 9/19/95) revised the Technical Specifications to remove the Makeup, Purification, and Chemical Addition systems from Section 3.2 of the Technical l
Specifications and the pertinent design information was relocated to the UFSAR. This proposed change servea tc conform the Technical Specifications to Amendment No.196 due to an administrative oversight in the corresponding TSCR No. 252. Therefore, the change is administrative in nature.
Markup TABLE 4.1-1 (Continued)
CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS
- 30. Borated Water Storage W NA F Tank LevelIrdicator
- 31. DELETED Be-i; Acid Mix Tr.h
- . In! CF=:! NA F ^. F b.T my :m:::: GF = ! M NA F
- 32. DELETED Ree! imed Beri: Acid Mix Tr.h
- . I=! Chr. :: NA NA F
- b. T .. 7 :::: C F = :! M NA F F. Technical Specifications Page 4-10:
Description Item 4 to Table 4.1-3 was revised to change sampling the spent fuel pool for boron concentration from after each makeup and monthly to weekly in accordance with the B&W Standard Technical Specifications (NUREG-1430) 3.7.15.1.
Item 6 to Table 4.1-3 was deleted to conform to Technical Specification amendment No.196, )
which deleted Section 3.2 and relocated the design information to the UFS AR based upon GPUN's TSCR no. 252, dated August i1,1995 that failed to request deletion of the corresponding surveillances ir. Section 4.1, which must also be deleted. Hence this TSCR serves to conform the Technical Specifications to amendment No.196.
Safety Evaluation l Item 4 to Table 4.1-3 was revised to weekly for the requirement to sample for boron concentration in the spent fuel pool in accordance with the B&W Standard Technical E
Am:ndm:nt No.196 due to an administrative oversight in the corresponding TSCR No.
252. Therefore, the change is administrative in nature.
. This propond change removes instrument surveillance requirements for the Boric Acid Mix Tank temperature instrument and the Reclaimed Boric Acid Storage Tank temperature instrument and is consistent with the Standard Technical Specifications for Babcock and
- Wilcox Plants, NUREG-1430, July 1992. Also this change meets the intent of the Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors issued July 1992 and codified in 10 CFR 50.36.
Safety Evaluation These items were changed to correct an administrative error. License Amendment No.
1% (dated 9/19/95) revised the Technical Specifications to remove the Makeup, Purification, and Chemical Addition systems from Section 3.2 of the Technical
- Specifications and the pertinent design information was relocated to the UFSAR. This proposed change serves to conform the Technical Specifications to Amendment No.196 due to an administrative oversight in the corresponding TSCR No. 252. Therefore, the change is administrative in nature.
Marklip TABLE 4.1-1 (Continued) {
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CHANNEL DESCRIlilON CHECK TEST CALIBRATE REMARKS
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- 30. Borated Water Storage W NA F Tank LevelIndicator
- 31. DELETED B -i; A:!d Mix Td
. 1 _. ._i cu _ _ _. ui ui e i E.y7. _E L.. ___.._..
. ,... Z.... i.'l. '
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- 32. DELETED Pa!:! :d Beri Acid Mix Td
- . L:v:! Ch: ::: "A NA F l
- b. T....,.n:n ; Ch .x:: " FA ;
F F. Technical Specifications Page 4-10: j Description Item 4 to Table 4.1-3 was revised to change sampling the spent fuel pool for boron concentration from after each makeup and monthly to weekly in accordance with the B&W Standard Technical Specifications (NUREG-1430) 3.7.15.1.
Item 6 to Table 4.1-3 was deleted to conform to Technical Specificition amendment No.196, which deleted Section 3.2 and relocated the design information to the UFS AR based upon GPUN's TSCR no. 252, dated August 11,1995 that failed to request deletion of the !
corresponding surveillances in Section 4.1, which must also be deleted. Hence this TSCR serves to conform the Technical Specifications to amendment No.196.
Safetv Evaluation Item 4 to Table 4.1-3 was revised to weekly for the requirement to sample for boron concentration in the spent fuel pool in accordance with the B&W Standard Technical
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' Specifications (NUREG-1430) 3.7.15.1. The criteria that govern the storage rack locations of fuel assemblies in the Spent Fuel Pool were developed without taking credit for boron. The analyses show that a boron concentration of 600 ppmb will meet the NRC maximum allowable reactivity value under the postulated fuel handling accident condition; if not moving fuel, no min'imum boron concentration is required. The spent fuel pool boron is normally about 2700 ppmb in order to match the refueling boron requirement in Technical Specification 3.8.4 in the reactor coolant system and the fuel transfer canal during refineling operations.
If water were added to the spent fuel pool, based upon the current high and low level alarm set points, it would not drop below 343' 6" and would not be increased above 345' - about a two foot change. In this pool with a maximum volume of about 435,000 gallons, the two-foot drop represents about 32,000 gallons, which would lower the pool to 403,000 gallons. Procedure N1800.2 (Chemistry Specifications) has an admmistrative limit of 2,650 to 5,000 ppm boron.
Using the lower limit of 2,650 ppm, the boron would drop to 2,455 ppm, which is well above the Technical Specification minimum value of 600 ppm. Even if the spent fuel pool boron dropped to as low as 650 ppm boron, the addition of 32,000 gallons of water would lower it to 602 ppm boron that is still above the minimum Technical Specification value. Therefore, routine makeups of the spent fuel pool with non-borated demineralized water would not reasonably be expected to exceed the minimum boron concentration limits.
Movements of fuel assemblies in the spent fuel pool storage racks are independently verified at the time of each movement to assure that the storage requirements of technical specification 5.4 are met. Makeup to the spent fuel pool is typically required due to evaporation of water from the surface of the pool, whici leaves the pool at a slightly higher born concentration. Makeup to the pool simply restores the pool level and boron concentration. These margins and the large volume of the spent fuel pool provide assurance that boron concentration in the pool will be maintained within the limit by a weekly boron concentration sampling frequency.
l Therefore, sampling weekly isjustified and is consistent with the RSTS. %c administrative changes have no affect upon safety.
Markup TABLE 4.1-3 Cont'd MINIMUM SAMPLING FREQUENCY 119m Qhesh Frecuency
- 4. Spent Fuel Pool Boron concentration greater than Weekly Water Sample - orequal to 600 ppmb M=&!y =d da each makeup G. Technical Specifications Page 4-41:
Description Tech Spec 4.5.2.1 requires that a high-pressure injection system performance test be performed each refueling interval. He current specification requires that testing must be done by an operator starting the pumps and a test signal must be used to open the valves. The requirement to use a " test signal" to open the valves adds complexity to the testing without adding value. Other Technical Specifications l
provide requitements to ensure that the high-pressure injection valve actuation logic is tested (TS 4.1-1 item .84) and the valve operation is tested (TS 4.2.2 & 4.5.2.4.a). These tests are performed quarterly using the ES logic test signal. During the refueling interval, the high-pressure injection performance test, adds complexity. The operators must carefully manipulate the ES test features and verify that only the desired components are actuated, at a time when their attention must also consider low-temperature over protection issues and verification of proper high-pressure injection component and system performance. Removing the ES test signal requirement would allow the valves to be opened as l
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'; part of the test setup and the risks associated with inadvertent ESAS actuation would be minimized.
This revision to 4.5.2.1 removes the requirement to use the test signal to open the valves and uses the measured flow to confirm system performance.
I Safety Ev'aluation
! De scope of testing of the Engineered Safeguards Actuation System (ESAS) and High-Pressure Injection (HPI) systems is not affected by this change. The change eliminates unnecessary overlap between requirements. The frequency of testing of the valve actuation logic and valve operation is not i affected. The required system performance is verified by system flow. System flow is the most l accurate and appropriate means to verify the capability of the system to meet accident analysis
! assumptions.
He change in test method does not degrade the quality or scope of the system performance evaluation.
De present test method requirement evaluates system performance based on the flow indication after l a " test signal" has been used to open the valves (MU-V-16 A/B or C/D). The test acceptance criteria l are unchanged. The HPI system configuration when HPI flow is measured is unchanged.
RSTS includec three separate requirements [3.5.2.4, 3.5.2.5,3.5.2.6] to (1) test pump performance IAW IST requuements (2) verify auto actuation of the HPI valves every 18 months and (3) verify the auto actuation of the pump every 18 months. The revised TMI Technical Specifications would include similar requirements except where the TMI requirements exceed the RSTS. The component operation on an ES actuation is tested quarterly vs. every 18 months per the Technical Specifications.
l Markup Specification ;
l 4.5.2.1 High Pressure Injection l l
- a. During each refueling interval and following maintenance or modification that
- afTects system flow characteristics, system pumps and system high point vents
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shall be vented, and a syste.a test shall be conducted to demonstrate that the I
system is operable.
l The makeup pump =d its required =ppo".ing n=iliariesrwiH be : arted ==ueuy by the Ope =ter =d a test sig=! v/i!! be applied ic the high pre =ure injection (HPI) valve MU V 16A/B/C/D to de=cn: tat :ctuation of the high pre =ure injec-tion
- y; tem for emerg=cy core eccling operation-
- b. The test will be considered satisfactory if the valves (MU-V-14A/B & 16A/B/C/D) .
l have completed their travel and the make-up pumps are running as evidenced by system flow the centrc! board compon=t operating 4ights. Minimum acceptable injection flow must be greater than or equal to 431 gpm per HPI pump when pump discharge pressure is 600 psig or greater (the pressure between the pump and flow l limiting device) and when the RCS pressure is equal to or less than 600 psig.
1 H. Technical Specifications page 4-42:
Descriotion Technical Specification 4.5.2.3 requires that a core flood system test be performed each refueling interval. The test verifies that the check valves between the core flood tanks and the reactor vessel will open as designed. %e present requirement specifies that the test be performed "while l depressurizing the reactor coolant system." Literal compliance with this phrase would not allow the test to be done when at cold shutdown or during plant heat up. The inservice test requirements L.- m--
r 1 I
(TS 4.2.2) for full flow testing of these same check valv:s is met by testing while the reactor coolant system is depressurized. The present version of 4.5.2.3 does not r.Ilow the inservice test (which verifies that the check valve opens and can pass the accident required flow rate) to be credited as a system test. The revision requires that the system test be completed, without
' specifying plant conditions during the test. A single test can then be perfor. ed to satisfy the l inservice test requirement and the system test requirement.
Safety Evaluation The core flood valves can be tested (open verification) with the RCS depressurized without impacting the quality of the test result. The test per specification 4.5.2.3 requires that operation of the check valves be verified by an indication that the core flood tank level has decreased (as
- described in the bases). His test, "Will water flow from the core flood tanks to the RV7" can be l performed with the RCS depressurized by opening the core flood tank isolation valves when core flood tank pressure is greater than RCS pressure.
Markup i
4.5.2.3 Core Floodina l a. During each refueling period, a system test shall be conducted to demonstrate l proper operation of the system. Dudeg ip.x;;d=:i= cf the R=::cr l Cec!=: Sy::=w Verification shall be made that the check and isolation l valves in the core cooling flooding tank discharge lines operate properly.
I. Technical Specifications page 4-43:
Description Technical Specification 4.5.3.1 (similar to item G above) requires that a system reactor building emergency cooling system performance test be perfe.med on a refueling interval. His specification also includes the same unnecessary complexity as found in 4.5.2.1. He specification requires that the valves be opened by a " test signal." This additional requirement does not add value to the test. The valve actuation logic (TS 4.1-1.18) and valve operation (TS 4.2.2 & TS - 4.5.3.2.a) is tested as required quarterly. This revision to 4.5.3.1 removes the requirement to use the test signal to open the valves and uses the measured flow to confirm system performance.
Safety Evaluation he scope of testing of the engineered safeguards actuation system and the reactor building emergency cooling system is not affected by this change. The change eliminates unnecessary overlap between requirements. The frequency of testing of the valve actuation logic and valve operation is not affected. He required system performance is verified by system flow. System flow is the most accurate and appropriate means to verify the capability of the system to meet accident analysis assumptions.
He change in test method does not degrade the quality or scope of the system performance j evaluation. %e present test method requirement evaluates system performance based on the component indications after a " test signal" has been used to open the valves. He revised test acceptance criteria uses flow to evaluate system performance. The RBEC system configuration when flow is measured is unchanged.
RSTS includes two related requirements [3.6.6.3,3.6.6.7] to (1) verify RBEC system flow capacity exceeds design every 31 days and (2) verify the auto actuation of the RBEC system every 18 months. He revised TMI Tech Specs would include similar requirements except that the -
component operation on an ES actuation is tested quarterly vs. every 18 months and the system flow would be verified on a refueling interval versus every 31 days.
IT 1 -
(TS 4.2.2) for full flow testing of these same check valves is met by testing while the reactor coolant system is depressurized The present version of 4.5.2.3 does not allow the inservice test (which verifies that the check valve opens and can pass the accident required flow rate) to be credited as a system test. The revision requires that the system test be completed, without specifying plant conditions during the test. A single test can then be performed to satisfy the inservice test requirement and the system test requirement.
Safety Evaluation ne core flood valves can be tested (open verification) with the RCS depressurized without impacting the quality of the test result. The test per specification 4.5.2.3 requires that operation of l the check valves be verified by an indication that the core flood tank level has decreased (as i
described in the bases). This test, "Will water flow from the core flood tanks to the RV7" can be performed with the RCS depressurized by opening the core flood tank isolation valves when core flood tank pressure is greater than RCS pressure.
Markup 4.5.2.3 Core Floodina
- a. During cau. refueling period, a system test shall be conducted to demonstrate proper operation of the system. Dud:;; aps=d= den ofi: R=:ter Cc !=: Sye:=, Verification shall be made that the check and isolation l
valves in the core cooling flooding tank discharge lines operate properly.
i I. Technical Specifications page 4-43:
Description Technical Specification 4.5.3.1 (similar to Item G above) requires that a system reactor building emergency cooling system performance test be performed on a refueling interval. His specification also includes the same unnecessary complexity as found in 4.5.2.1. The specification requires that the valves be opened by a " test signal." This additional requirement does not add value to the test. The valve actuation logic (TS 4.1-1.18) and valve operation (TS 4.2.2 & TS 1 4.5.3.2.a) is tested as required quarterly. This revision to 4.5.3.1 removes the requirement to use the test signal to open the valves and uses the measured flow to confirm system performance.
Safety Evaluation De scope of testing of the engineered safeguards actuation system and the reactor building emergency cooling system is not affected by this change. %e change eliminates unnecessary l
overlap between requirements. The frequency of testing of the valve actuation logic and valve
{
operation is not affected. He required system performance is verified by system flow. System
^
flow is the most accurate and appropriate means to verify the capability of the system to meet accident analysis assumptions.
De change in test method does not degrade the quality or scope of the system performance evaluation. .De present test method requirement evaluates system performance based on the component indications after a " test signal" has been used to open the valves. The revised test acceptance criteria uses flow to evaluate system performance. De RBEC system configuration i when flow is measured is unchanged. i RSTS includes two related requirements [3.6.6.3, 3.6.6.7] to (1) verify RBEC system flow capacity exceeds design every 31 days and (2) verify the auto actuation of the RBEC system every 18 months. The revised TM1 Tech Specs would include similar requirements except that the component operation on an ES actuation is tested quarterly vs. every 18 months and the system flow would be verified on a refueling interval versus every 31 days.
- 3. Operation of the facility in accorduce with the proposed amendment will not involve a significant reduction in a margin of safety because no changes to plant operating limits or limiting safety system settings are proposed.
IU IMPLEMENTATION GPU Nuclear requests that the amendment authorizing these changes become effective within 60 days ofissuance.
I 1
j
~, 'y 4.5.3.1 System Tests
, a. Reactor Buildina Sorav System
- 1. At each refueling interval and simultaneously with the test of the emergency loading sequence, a reactor building 30 psi high pressure test signal will start the spray pump.
Except for the spray pump suction valves, all engineered safeguards spray valves will be closed.
Water will be circulated from the borated water storage tank through the reactor building spray pumps and returned through the test line to the borated water storage i tank.
1
( he operation of the spray valves will be verified during the component test of the R. B.
cooling and isolation system.
l he test will be considered satisfactory if the spray pumps have been successfully started as evidenced by the control board component operating lights, and either the
! station computer or pressure / flow indication.
- 2. Compressed air will be introduced into the spray headers to verify each spray nozzle is l unobstructed at least every ten years.
t I
l
- b. Reactor Buildina Coolina and Isolation Systems 1.During each refueling period, a system test shall be conducted to demonstrate proper operation of the system. A :::: :!;=! . "! =:=:: 6: R=c:= B !! ding E==g =y C !!:g :y:':: =!v ::: d:=:=:=:: ep= bili:y of th ::!=:.
l
- 2. De test will be considered satisfactory ific =!ve: h:ve ==p!; :d $ !r =p=::d := vel i = :vid==d by $ :=:=! bord ::=p = : p= ting !!;h:: =d ==d === cf
! veriE=:!=, =:h =: Se :tti= :=p : , 'c=! veifi=:i=, ve-ifi=ti= cf p==re/f!:w,
[ = =..:.2! he=d ec.npc= : p;mting light: i.":!::: by =;==:: !" :veitch :=tet:.
l Measured system flow is greater.than accident design flow rate.
1.
III NO SIGNIFICANT HAZARDS CONSIDERATIONS GPU Nuclear has determined that the requested Technical Specification Change poses no significant hazard consideration as defined byl0 CFR 50.92.
- 1. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probdility of occurrence or consequences of an accident previously evaluated. The pawcd amendment makes administrative corrections, adds l conditions to limiting condition', for operation, revises selected time clocks and l surveillance requirements consistent with NUREG 1430, and adds a time clock to a
! unique LCO. These changes have no affect upon the plant design or operation. The reliability of systems and components relied upon to prevent or mitigate the consequences of accidents previously evaluated is not degraded by the proposed changes.
Berefore, operation in accordance with the proposed amendment does not involve a significant increase in the probability of occurrence or consequences of an accident i previously evaluated.
- 2. Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any previously evaluated, because no new accident initiators would be created.
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