ML20044G796
| ML20044G796 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/26/1993 |
| From: | Shelton D CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| REF-GTECI-070, REF-GTECI-094, REF-GTECI-NI, TASK-070, TASK-094, TASK-70, TASK-94, TASK-OR 2128, GL-90-06, GL-90-6, NUDOCS 9306040281 | |
| Download: ML20044G796 (29) | |
Text
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fm W _ CENTERf0R ENERGY l
Donald C. Shelton 300 Madison Avenue Vice President Nuclear Toledo, OH 436520001 Davis Besse (419)249 2300 Docket Number 50-346 i
.i License Number NPF-3 ~
Serial Number 2128 May 26, 1993 United States Nuclear Regulatory Commission ,
Document Control Desk +
Vashington, DC 20555
Subject:
Generic Letter 90-06, Resolution of Generic Issue 70, Pover-0perated Relief Valve and Block Valve Reliability, and. ,
Generic Issue 94, Additional Lov-Temperature Overpressure '
Protection for Light-Vater Reactors i Gentlemen:
Nuclear Regulatory Commission (NRC) Generic Letter (GL) 90-06 (Toledo :
Edison letter Log Number 3267), dated June 25, 1990, provided~the NRC .]"
staff's positions resulting from the resolution of Generic Issue 70.'
(GI-70), Power-Operated Relief Valve (PORV) and Block Valve Reliability, and Generic Issue 94, Additional Low-Temperature j Overpressure Protection for Light-Vater Reactors. NUREG-1316, Technical Findings and Regulatory Analysis Related to Generic Issue 70,-
was included as Enclosure C to GL 90-06. As stated in GL 90_-06, '
Generic Issue 70 is applicable to Babcock and.Vilcox (B&V)-designed. l plants, whereas Generic Issue 94 does not apply to B&V-designed plants.
- such as the Davis-Besse Nuclear Power Station (DBNPS). Therefore, GI-94 is not addressed in this response. 'i I
Generic Letter 90-06 requested that Toledo Edison respond to three '
items regarding GI-70 for the DBNPS: ;
Whether Toledo Edison vould include the Power-Operated Relief Valve (PORV) and Block Valve within its Nuclear-Quality Assurance (0A)- :
Program, implement an appropriate. maintenance and maintenance-training program for the PORV and its block valve, and procure spare and replacement parts (as well as complete components)'in ;
accordance with original construction codes and standards, j 9306040281 930526-
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Operahng Companies: A f l
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( Docket Number 50-346 2
License Number NPF-3 ;
% Serial Number 2128 Page 2
- 2. Whether Toledo Edison would include the PORV and block valve within-its'American Society of Mechanical Engineers-(ASME) Boiler and Pressure Vessel Code Pump and Valve-Testing Program, include the PORV's block valve in the licensee's expanded MOV test program discussed in NRC GL'89-10 (Safety-Related Motor Operated Valve .
Testing and Surveillance), and l i
- 3. Whether TE vould commit to converting to the model Technical i Specifications on the PORV and block valve contained within ;
GL 90-06 (Attachment A-4 of Enclosure A) for B&W-designed plants.
By letters dated December 21, 1990 (Toledo Edison Serial Letter Number 1884) and June 26, 1992 (Toledo Edison' Serial Letter ;
Number 2046), Toledo Edison responded to GL 90-06. Toledo Edison was ;
in general agreement with items 1 and 2 above (NRC Recommendations j 3.1.1 and 3.1.2). In Serial Letter Number 1884, Toledo Edison j concluded that the DBNPS Nuclear Quality Assurance program, Pump and Valve Testing Program, and its maintenance and maintenance training j program vill ensure reliable PORV and block valve operation. .However, Toledo Edison took an alternative position with regards to item 3 (NRC- ,
Recommendation 3.1.3) and provided its justification for not committing to the use of the model Technical Specifications of GL 90-06. Toledo :
Edison's position on Recommendation 3.1.3 (model Technical ,
Specif ations) concluded that the existing DBNPS Technical i Specifications for the PORV and block valve contain appropriate '
requirements for their design and operation, and that Toledo Edison should not utilize the NRC's generic model Technical Specifications of Attachment A-4 in Enclosure A of GL 90-06. However, a Technical ~ :
Specification change to Bases 3/4.4.11 was submitted to the NRC by '
Toledo Edison Serial Letter Number 2046 to add more discussion of the :
functions of the PORV and block valve in the DBNPS system configuration. i Nuclear Regulatory Commission (NRC) letter (Toledo Edison Letter Log i Number 3850), dated October 14, 1992, provided the NRC staff's reviev' }
of Toledo Edison's response to GL 90-06. The NRC staff's review of ;
Toledo Edison's submittals (Toledo Edison Serial Letters Number 1884 ,
and Number 2046) stated that Toledo Edison had not responded to !
selected portions of GL 90-06. Specifically, the NRC staff identified !
the following two issues- .;
- 1. The NRC staff stated regarding the NRC recommendation for PORV.
stroke testing in Modes 3 or 4: "Your submittal did not adequately meet this staff position. The staff is not accepting Mode 5 (Cold i Shutdown) testing simply because it is allovable by the ASME Code t or that the NRC-approved IST program includes Mode 5 for this particular test." i i
- 2. The NRC staff also stated: "The staff position for the resolution of GI-70 required TS upgrades is as presented in the generic j letter. Our review of your submittal indicates that you have ('
declined to respond to the staff position."
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$- Docket Number 50-346 I License Number NPF-3
, !. Serial Number 2128 Page 3
- Therefore, Toledo Edison was requested to resubmit a response to address these'two NRC issues.
Toledo' Edison believes the NRC staff's letter of October 14, 1992, I mischaracterized Toledo Edison's responses of December 21, 1990 and June 26, 1992. Regarding the first issue (NRC recommendation for PORV '
stroke testing in Modes 3 or 4), Toledo Edison did not take the position that Mode 5 testing was acceptable simply because it is allowed by the ASME Code and that the NRC-approved Pump and Valve Inservice Testing program includes Mode 5 for this particular test.
Rather, Toledo Edison based its position on the constraints of limiting any Reactor Coolant System pressure transient and the generation of radioactive vaste, and stated this as the basis in its letter of December 21, 1990.
Regarding the second issue (NRC recommendation for Technical Specification changes), Toledo Edison did not decline to respond to the i NRC staff's position in GL 90-06. Rather, Toledo Edison disagreed with ;
the NRC staf f's recommendation and provided technical justification for its position in the December 21, 1990 letter. The justification .i concluded that the existing Technical Specifications contain ;
appropriate requirements for the DBNPS, and that adoption of the ,
NRC-recommended generic Technical Specifications in their entirety is !
inappropriate. Toledo Edison does not believe that safety would be i improved since an increase in induced plant transients from plant !
shutdowns could result from the incorporation of the Action statements
- of GL 90-06 model Technical Specifications.
As a result of the tac staff having issued or plans for issuing similar letters (as that sent to Toledo Edison dated October 14, 1992) to the other B&W Owners Group (BV0G) utilities with operating plants, a BV0G initiative was formed with Toledo Edison's participation. This BV0G 3
initiative is appropriate because the application of new model ,
i Technical Specifications is a generic-issue and of concern to each of ;
the BV0G utilities. Toledo Edison's Serial Letter Number 2102, dated December 15, 1992, to the imC discussed this initiative.
A telephone conference call was held between the IEC staff and BV0G ;
representatives on December 17, 1992, to discuss the GL 90-06 recommendation for Technical Specifications. A BV0G letter.(Number OG-1128), dated January 18, 1993, was submitted to the NRC providing [
justification for the BV0G's position that the model Technical !
Specifications are inappropriate. In addition to the BWOG 1etter, i' attached is Toledo Edison's response to the NRC's October 14, 1992 letter, which further addresses the NRC's recommendation for Technical Specifications changes and also addresses the istue of PORV stroke testing in Modes 3 or 4. This response provides further justification as to the appropriateness of the existing Technical Specifications rather than the model Technical Specifications, to the DBNPS system design and operation.
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.' . Docket' Number.50-346 -!
I License Number NPF-3
!. Serial Number 2128 Page 4 1
Should you have any questions concerning these matters, please contact .l Mr. R. V. Schrauder, Manager - Nuclear Licensing at (419) 249-2366.- i Very tr '..y yours, l 1 1 ___
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I RAS /dl Attachments cc: S. Stasek, NRC Senior Resident Inspector A. B. Davis, Regional Administrator, NRC Region III J. B. Hopkins, NRC/NRR DB-1 Senior Project Manager Utility Radiological Safety' Board l
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'5 Docket Number 50-3467 s-License Number NPF-3 Serial Number 2128
!. Enclosure-lPage 1 RESPONSE TO GENERIC LETTER 90 l FOR
'l DAVIS-BESSE NUCLEAR POWER STATION.. l
.l l
UNIT NUMBER 1 This letter is submitted pursuant to 10CFR50.54(f). Enclosed is Toledo 'l Edison's response (letter Serial Number 2128) to the NRC's letter dated- l October 14, 1992, regarding Generic. Letter 90-06, Resolution of Generic-Issue 70, "Pover Operated Relief Valve and Block Valve Reliability,"
and Generic Issue 94, " Additional Low-Temperature Overpressure-Protection for Light-Vater Reactors."
j '
By: \i b /\ V _
D. C.'Shelt6n,'Vice President - Nuclear.
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Sworn and subscribed before me this 26th day of May,1993.
11/bhY l 10A Notary P)/b1'ic, 5thfe 'of Ohio EVELYNLORESS NODWFReuC.M4OFCHO
%hEghsME,MM
-' Docket Number 50-346
. License Number NPF-3 Serial Number 2128 Attachment 1 Page 1 GENERIC LETTER.90-06 TOLEDO EDISON RESPONSE TO NRC STAFF.REVIEV LETTER I. NRC RECOMMENDATION ITEM 3.1.2:
Include PORVs, valves in PORV control air systems, and block valves within the scope of a program covered by subsection ~IVV,.
" Inservice Testing of Valves in Nuclear Pover Plants," of Section XI of the ASME Boiler and Pressure Vessel Code. Stroke testing of PORVs should only be performed during Mode 3 (Hot Standby) or Mode 4 (Hot Shutdown) and in all cases prior to establishing conditions where the PORVs are used for low-temperature overpressure protection. ' Stroke testing of the.
PORVs should not be performed during power operation.
Additionally, the PORV block valves should be included in the licensees' expanded MOV test program discussed in NRC GL 89-10,
" Safety-Related Motor Operated Valve Testing and Surveillance,"
dated June 28, 1989.
NRC OCTOBER 14, 1992 LETTER POSITION:
The staff maintains its position that the.PORVs should be stroke tested in Modes 3 or 4 in order to verify.the capability to function in an environment more representative of operating conditions. .In your revised response, discuss how PORV stroke testing provides assurance that the PORVs will perform all necessary safety functions adequately at the required system operating conditions.
TOLEDO EDISON RESPONSE TO RECOMMENDATION ITEM'3.1.2:
Toledo Edison's response to Recommendation Item 3.1.2 remains the same as previously discussed in Toledo Edison Serial Letter 1884 (dated December 21, 1990), except concerning the testing of the PORV during Mode 3 (Hot Standby) or Mode ~4 (Hot Shutdown).
Toledo Edison tests the PORV in Mode 3 (Hot Standby) during cooldown to or heatup from Mode 5 (Cold Shutdown) in accordance with procedure DB-SP-03363 (Pressurizer Power Operated Relief Valve Cycle Test) if the PORV has not'been cycled within the previous three months. Procedure DB-SP-03363 verifies the PORV will fully open and then close when its associated solenoid is cycled (the FORV is a solenoid actuated, pilot operated relief valve and, as such, control air is not used to actuate this valve). The PORV disc movement is confirmed by exercising the valve while observing the PORV solenoid position and flow indicators. As discussed before, the PORV-is cycled and not stroke tested in Mode 3 because stroke testing. requires that a timed pressure drop must be observed over several minutes which could potentially lead to a RCS system pressure transient and the generation of radioactive vaste.
As previously discussed in Toledo Edison Serial Letter 1884 (dated December 21, 1990), the PORV is stroke tested at
Docket Number 50-346 License Number NPF-3 [
Serial Number 2128 ;
Attachment 1 Page 2 refueling intervals while the plant is in Mode 5 per Technical i
Specification (TS) 4.0.5 and procedure DB-SP-03366 (Reactor [
Coolant System Vent Path Operability). Procedure DB-SP-03366 !
' also verifies that the PORV flovpath is capable of passing flow .
per TS 4.4.11 requirements. Since the PORV does not have direct position indication, the amount of opening (valve lift) must be j inferred. This inference is based on a timed measurement of the ,
pressurizer pressure decrease. This procedure measures an ;
acceptable flow through the PORV by timing a pressure drop from l approximately 125 psig to 115 psig. This 10 psig pressure drop l is then correlated to flow and must occur within a set time ;
period (e.g., RCS pressure should drop a minimum of 10 psig I within 5 minutes). This pressure drop test over a specified ;
period of time is also used for trending degradation of the PORV.
Reliability testing has also been performed on the DBNPS's Crosby ,
style HPV-SN PORVs. The Crosby style HPV-SN PORV was tested in [
the EPRI Safety and Relief Valve Test Program (reference !
NUREG-0737, ltem II.D.1, and Toledo Edison letter Serial [
Number 905, dated February 1, 1983) and was found to perform ;
vithout failure to open or close on demand under steam and water '!
flow conditions. Crosby style HPV-SN PORVs were further tested by Toledo Edison at the Duke Power Marshall Station in October !
i and November of 1985 (reference Toledo Edison letter Serial' ^
Number 1308, dated October 14, 1986). During the October 1985
! PORV Operability Test, the Crosby HPV-SN style valve was ;
I successfully opened and closed a total of 100 times at'two l different back pressures (50 valve cycles at approximately !
425 psig and 50 valve cycles at approximately 160 psig). i Subsequent to the 1985 October PORV Test, a modified nozzle and ,
disc vere installed in a Crosby style HPV-SN PORV to increase the [
valve's flow rate. The modified Crosby style HPV-SN PORV vas j then flow tested at the Marshall Station in November 1985. l During the November 1985 PORV Test, the modified valve !
i successfully opened and closed a total of 20 times at a backpressure of approximately 500 psig. A modified HPV-SN PORV 4
was installed in the DBNPS, and confirmation of its operability ,
was successfully accomplished during the 1986 plant startup. l I
During the 1986 startup, the modified PORV was successfully opened and closed a total of seven times at a system pressure of :
I approximately 700 psig (Mode 3). This modified PORV was also successfully opened and closed, a total of three_ times (with no l failures) at a full RCS pressure of approximately 2155 psig. j Since this past testing at both high and low pressure / temperature ;
has been performed on Crosby style HPV-SN PORVs without failure, f the present testing requirements of.the PORV to meet TS 4.4.11 and TS 4.0.5, in conjunction with the procedurally - required !
cycle test (procedure DB-SP-03363) of the PORV in Mode 3, j represents the PORV's ability to reliably perform its function. i Therefore, the present PORV testing provides assurance of PORV !
l operability for the DBNPS. !
- - Docket Number 50-346 :
License Number NPF-3 l Serial Number 2128 !
- . Attachment 1 Page 3 The recommendation to stroke test the PORV in all cases prior.to entering the low-temperature overpressure protection (LTOP) mode l is an LTOP - driven recommendation directly related to the j necessity of having a PORV available for LTOP. Applying-this ;
recommendation to B&V-designed plants such as the DBNPS is inconsistent with GL 90-06. Generic Letter 90-06 specifically ;
states LTOP recommendations of the letter (i.e., the resolution 1 of Generic Issue 94) are not applicable to B&V-designed plants. >
Furthermore, the PORV is not required for LTOP at the DBNPS since LTOP relief is provided by the Decay Heat Removal System dropline relief valve DH-4849. The NRC staff's October 14, 1992 letter states that Generic Issue 94 is not applicable to the DBNPS.
As previously discussed in Toledo Edison Serial Letter 1884 !
(dated December 21, 1990), the block valve, RC11, is' tested in i accordance with the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IVV, 1986 Edition. The block valve and !
its motor operator are included in the DBNPS Motor-Operated Valve ?
Reliability and Improvement Program under NRC GL 89-10 as-requested by GL 90-06. Accordingly, Toledo Edison has no differences with NRC Recommendation Item 3.1.2 regarding the block valve. i i
II. NRC RECOMMENDATION ITEM 3.1.3:
For operating PVR plants, modify the Limiting Conditions for Operation of PORVs and block valves in the TS for Modes 1, 2,:and 3 to incorporate the position adopted by the NRC staff in recent ,
licensing actions. Attachments A-4 and A-5 (to Generic Letter >
90-06) are provided for guidance. !
NRC OCTOBER 14, 1992 LETTER POSITION: .
t The NRC staff vill not accept, without sufficient justification, .
the position that the TS upgrades are unnecessary because the !
PORVs are not the primary means of dealing with the three safety j functions identified in the generic letter.
- TOLEDO EDISON RESPONSE TO RECOMMENDATION ITEM 3.1.3
- ;
Vith regards to the NRC staff's position provided in the NRC l letter dated October 14, 1992 (Toledo Edison Log Number 3850),
Toledo Edison's response to Recommendation Item 3.1.3 is !
augmented with additional information to clarify and further justify Toledo Edison's original position on this recommendation. -
A. NRC MODEL AND DBNPS TECHNICAL SPECIFICATIONS COMPARISON: ,
The DBNPS has a Babcock and Wilcox (B&V)-designed Nuclear +
Steam Supply System which contains one PORV and one block valve. Two ASME Boiler and Pressure Vessel Code safety valves (RC13A and RC13B) and the PORV (RC2A) are connected to l l
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Docket Number 50-346 l License Number NPF-3 Serial. Number 2128 l l Attachment 1 ;
Page 4 l
the top of the pressurizer, and provide overpressure ;
protection for the RCS. Each of the Code safety valves i discharges to the containment atmosphere. The PORV j discharges to the pressurizer quench tank. Another )
pressurizer vent path for the RCS is provided through two !
motor-operated valves (RC239A and RC200) to the pressurizer quench tank. The pressurizer establishes and maintains RCS -]
pressure within prescribed limits and provides a surge ;
chamber and a water reserve to accommodate changes in reactor .i coolant volume during operation. The RCS pressure is also j decreased by injection of reactor coolant from the Reactor !
Coolant Pump (RCP) 2-2 discharge via the pressurizer spray ,
nozzle. The RCS pressure is increased by energizing . .
immersion type pressurizer heaters. The motor-operated PORV .
block valve (RC11) provides the capability to remotely !
isolate the PORV in the event that the PORV fails open, fails !
to reseat, or is leaking.
The following discussion compares the recommended GL 90-06 :
model TS (Enclosure A, Attachment A-4) with the present.DBNPS TS, their suitability to the DBNPS design and operation, and !
to existing DBNPS procedural requirements. l I
.The GL 90-06 model TS contain an Action requirement (3.4.4 Action a) that in case of excessive PORV seat leakage, the s PORV be restored or the block valve be closed within the hour with the electrical power maintained on. Should this not be accomplished, then the plant must be in Mode 3 in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 4 in the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Davis-Besse i Nuclear Power Station TS 3/4.4.11, RCS-RCS Vents, contains requirements on the PORV and block valve that are applicable ..
- during Modes 1, 2, and 3. For the pressurizer, there are,two i possible vent paths (besides the two pressurizer code safety valves)
- one through valves RC11 (PORV block valve) and RC2A ,
(PORV) and one through motor-operated valves RC239A and l j~ RC200 (pressurizer vent line). Therefore, under the DBNPS [
design should the block valve be closed due to excessive FORV leakage, a second pressurizer vent path remains available for i the steam generator tube rupture (SGTR) event and plant I natural circulation cooldowns, and the continued availability i of electrical power to the block valve is less of a concern.
! Furthermore, the DBNPS TS do not require the removal of l
. electrical power from a block valve closed due to excessive i PORV leak. age, and the stated concern of NUREG-1316 and GL i 90-06 that some plants' TS do specifically require removal of ;
power is addressed. As discussed later in Section B of this i 2- response, there is no credit taken in the DBNPS Updated j Safety Analysis Report for the usage of the PORV and block
- valve vent path or pressurizer vent line for accomplishing
- RCS depressurization during a SGTR, and the calculated doses are vell within the 10CFR Part 100 guidelines. }
Docket Number 50-346 l
, License Number NPF-3 Serial Number 2128 l Attachment 1 l Page 5 I i
Technical Specification 3/4.4.11 also provides TS j requirements for the RCS vent paths. There are three RCS i vent paths: 1) one high point vent is installed at.the top of l the RCS Loop 1 hot leg, 2) another high point vent is installed at the top of the hot leg in RCS Loop 2,-and 3) ;
a third RCS vent path on the pressurizer through either the !
PORV flovpath (RC2A and RC11) or the pressurizer vent line. l (RC239A and RC200). Therefore, there are three vent paths to +
promote proper cooling of the core. By requiring operability l of these vents, steam or non-condensible gas bubbles can be !
eliminated following the loss of natural circulation. These !
RCS vents can be used to restore natural circulation !
following a small break loss-of-coolant-accident or to enable i system cooldown following an inadequate core cooling event. '
Technical Specification 3.4.11 presently requires at-least l one vent path to be operable at each location during Modes 1, ;
2, and 3. If there is no vent path operable at anyone of the 'l three locations, then the inoperable vent path location is !
required by TS 3.4.11 to be restored to operable status '
within 30 days, or the plant placed in Hot Standby (Mode 3) ;
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Hot Shutdown (Mode'4) within the following. !
30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. If there is no vent path operable in two of the i three locations, then at least one.of the inoperable vent path locations must be restored.to operable status within i 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the plant placed in Hot Standby (Mode 3) within 6 ;
hours and in Hot Shutdown (Mode 4) within the following 30' ,
hours. With all three of.the vent path locations inoperable, !
two of the inoperable vent path locations must be restored to ;
operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or the plant placed in Hot ?
Standby (Mode 3) within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Hot Shutdown (Mode 4) .
within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. l In the event of requiring feed and bleed cooling, under the DBNPS design, the combination of two pressurizer code safety ;
valves and two makeup pumps can also provide the necessary i j supporting Reactor Coolant System (RCS) configuration. l
' Technical Specification 3.1.2.4 requires that two makeup {
pumps be operable in Modes 1 through 4 (with RCS pressure l greater than or equal to 150 psig) and TS 3.4.3 requires both l pressurizer code safety valves to be operable in Modes 1 ;
through 3. !
Therefore, since: 1) the existing DBNPS TS already do not !
require removal of electrical power from the block valve should it be closed due to excessive PORV seat leakage, 2) an i alternate TS-required RCS vent path, in addition to the PORV l and block valve path, exists to respond to a SGTR or to i restore natural circulation, and 3) feed and bleed cooling can be accomplished without the PORV and block valve path being available by using other plant equipment covered by TS, the Generic Letter 90-06 model TS 3.4.4 Action a recommendation (which does not take into account the DBNPS 1 multiple vent path design) is not appropriate for the DBNPS. i 4 i I
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Docket Number 50-346
. License Number NPP-3 i Serial Number 2128 ;
Attachment 1
- Page 6 ,
i GL 90-06, TS 3.4.4, ACTION B ,
The GL 90-06 model TS contain an Action statement (3.4.4 )
Action b) that requires if the FORV is inoperable, for reasons other than excessive seat leakage, that.the PORV be restored to operable status or the block valve be closed '
within one hour with its power removed. If the block valve is closed, then the plant is to be placed in Mode 3 in the next six hours and in Mode 4 in-the following six hours. ,
Existing DBNPS TS 3.4.3 requires that if the PORV becomes ;
inoperable it: 1) either be isolated, or 2) TS 3.0.3 entered ;
with action initiated within one-hour to shutdown the plant, !
in at least Mode 3 vithin the next six hours and in at least ,
Mode 4 within the following six hours. As discussed above, ,
the existing DBNPS TS 3.4.11, 3.1.2.4, and 3.4.3 appropriately take into consideration the DBNPS's diverse Reactor Coolant System vent paths available for recovery from :
a SGTR event, a plant natural circulation cooldown, and the j use of alternative means for promoting feed and bleed 4 cooling, should the PORV and block valve path become i inoperable. 'Furthermore, as discussed in Section B of this response, no credit is taken in the DBNPS Updated Safety Analysis Report for the usage of the PORV and block valve ;
vent path or the pressurizer vent line for accomplishing RCS ;
depressurization during a SGTR, and the calculated doses are l vell within the 10 CFR Part 100 guidelines. Therefore, GL.
90-06 model TS 3.4.4 Action b does not appropriately account !
for the DBNPS design. .i GL 90-06, TS 3.4.4, ACTION C r The GL 90-06 model TS contain an Action statement (3.4.4 l Action c) that requires if the block valve is inoperable, within one hour restore the block valve to operable status or ;
the PORV must be placed in manual control and the block valve j i restored in the next hour, or the plant placed in Mode 3 !
' within the next six hours and in Mode 4 vithin the following ?
l six hours. As discussed above, the present DBNPS TS appropriately take into consideration the DBNPS's diverse
, Reactor Coolant System vent paths for recovery from a SGTR -
event, a plant natural circulation cooldown, and the alternative means for promoting feed and bleed cooling, 4 should the PORV and block valve path become inoperable.
Furthermore, as discussed in Section B of this response, no credit is taken in the DBNPS Updated Safety Analysis Report for the usage of the PORV and block valve vent path or the i pressurizer vent line for accomplishing RCS depressurization during a SGTR, and the calculated doses are well within the j
- 10 CFR Part 100 guidelines. Therefore, GL 90-06 model TS ,
- 3.4.4 Action c does not appropriately account for the DBNPS +
j design. ,
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Docket Number 50-346 j
+
, License Number NPF-3 *
-' Serial Number 2128' !
Attachment 1 !
Page 7 j GL 90-06, TS SURVEILLANCE REQUIREMENT 4.4.4.1.A The GL 90-06 model TS contain a surveillance _ requirement (4.4.4.1.a) to operate the PORV through one complete cycle of full travel during Modes 3 or 4 at least once per-18 months.
As previously discussed in response to NRC Recommendation l Item 3.1.2, the DBNPS PORV is cycled through full travel in :
4 Mode 3 (Hot Standby) during the cooldown to or the subsequent heatup from Mode 5 if the PORV has not been cycled within the i
~
previous three months. This cycling of the'PORV (at a frequency of approximately 18-22.5 months) is in accordance 1 with procedure DB-SP-03363 (Pressurizer Pover' Operated Relief Valve Cycle Test) which verifies the PORV will. fully open and {
close when its associated solenoid is cycled. The testing i has been performed in the past and is presently being performed without the testing being required by a TS. i NUREG-1316 estimates that the cost to implement the GL 90 model TS is $16,000. Since the testing is already being ,
accomplished, to add it to the TS vould only result in an _
j expenditure of Toledo Edison resources without any comparable !
benefit in safety. Therefore, Toledo Edison is not proposing _
the incorporation of this model TS surveillance requirement. !
GL 90-06, TS SURVEILLANCE REQUIREMENT 4.4.4.1.B '
The GL 90-06 model TS (applicable in Modes 1, 2 and 3) :
contain a surveillance requirement (4.4.4.1.b) to perform a- :
channel calibration of the PORV actuation instrumentation at I i least once every 18 months. Davis-Besse Nuclear Power l Station TS 3/4.4.3, RCS-Safety Valves and Pilot Operated i Relief Valve - Operating, (applicable in Modes 1, 2 and 3) ;
I contains requirements for the PORV. Technical Specification !
4.4.3 requires that a channel calibration check be performed j l every 18 months for the PORV. Therefore, this NRC ;
j recommendation is already addressed in the DBNPS TS. l GL 90-06, TS SURVEILLANCE REQUIREMENT 4.4.4.2 The GL 90-06 model TS contain a surveillance requirement *
(4.4.4.2) to cycle the block valve every 92 days. The DBNPS ,
block valve is cycled through a complete cycle in accordance with TS 4.0.5 and the Inservice Testing Program every 92 ;
days. Therefore, since this testing is already being !
accomplished in accordance with TS 4.0.5, to add it as a duplicative requirement vould only result in an expenditure of Toledo Edison and NRC staff resources without any ,
comparable benefit in safety. Therefore, Toledo Edison is 'l not proposing the duplicative incorporation of this model TS surveillance requirement.
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GL 90-06, TS SURVEILLANCE REQUIREMENT 4.4.4.3 l
The GL 90-06 model TS contain an 18-month surveillance ,
requirement (4.4.4.3) to demonstrate that emergency power can (
be transferred to the PORV and block valve motive'and control j circuits and then the PORV and block valve operated through a l complete cycle. During the DBNPS Fifth Refueling Outage, the l PORV power source was reconfigured to be supplied from an i essential DC Distribution Panel, D2N, and the solenoid coil on the PORV was environmentally qualified for the containment ;
atmosphere as calculated for the feed and bleed scenario. !
During the DBNPS Sixth Refueling Outage, the circuitry ;
required to manually open the PORV was upgraded to Class IE.
The installed upgrade vas designed to account for voltage degradation that may occur after postulated battery use. !
Also, the power supply associated with the PORV demand position indicating lights was transferred to an essential-120 AC instrument distribution panel. The block valve is povered from 480v vital bus E16B. Vital bus E16B can be j supplied by an emergency diesel generator. This strveillance requirement to demonstrate that emergency power can be i 4 provided (and is performed by transferring from.the normal i
supply.to the emergency supply with valve cycling) is not j necessary because the normal power supplies for the PORV and block valve already are Class IE essential power supplies at the DBNPS. Per the GL 90-06 model TS Bases (3/4.4.4), this i surveillance requirement is provided for those plants where the PORV and block valve are installed with non-safety-grade power sources (motive and control pover) and provided with t backup (emergency) power sources. Therefore, this model TS !
4 surveillance requirement is not appropriate to the DBNPS i
]
design since Class 1E essential power already is the normal i power supply.
TS REQUIREMENTS BEYOND GL 90-06 l
[
The DBNPS TS also contain the following requirements beyond ;
i that recommended by GL 90-06. )
i Technical Specification 4.4.11 requires that each RCS vent path (including that through the PORV and block valve)'to be '
demonstrated operable at least once per 18 months by:
- 1. Verifying all manual isolation valves in each vent path are locked in the open position, and
- 2. Cycling each valve in the vent path through at least one complete cycle of full travel from the control room during Cold Shutdown (Mode 5) or Refueling (Mode 6), and j 3. Verifying flow through the Reactor Coolant System vent paths during Cold Shutdown (Mode 5) or Refueling (Mode 4
6).
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' Docket Number 50-346
- License Number NPF-3 I Serial Number 2128 Attachment'1 Page 9 It is important to note that the vent path through the PORV and block valve is an alternate method of removing non-condensible gases from the pressurizer (i.e., the PORV and block valve path does not need to be operable as a RCS vent per TS 3.4.11, if the pressurizer high point vent line through RC239A and-RC200 is operable).
The GL'90-06 model TS vould allow plant startup while relying on an Action statement (i.e., TS 3.0.4 is not applicable).
However, DBNPS TS 3.0.4 is applicable to TS 3/4.1.2.4 regarding the Makeup pumps and to TS 3/4.4.3 regarding the pressurizer code safety valves.
LOV-TEMPERATURE OVERPRESSURE PROTECTION l
The PORV is not required for low-temperature overpressure protection (LTOP) at the DBNPS. Low-temperature Overpressure protection relief is provided by the Decay Heat Removal System dropline relief valve, DH-4849, sized to provide l low-temperature overpressure protection during Modes 4 and 5.
The Limiting Condition for Operation (LCO) for DH-4849 for LTOP considerations is provided by DBNPS TS 3/4.4.2, Safety Valves - Shutdown. Therefore, LCOs for the PORV and block valve are not appropriate for LTOP considerations in Modes 4 and 5.
SUMMARY
In summary, the GL 90-06 model TS recommendations for a channel calibration of the PORV actuation instrumentation and cycling of the block valve every 92 days are included in existing DBNPS TS. The GL 90-06 model TS recommendation for a periodic demonstration of the transfer of emergency power to the PORV and block valve motive and control circuits is not necessary since Class lE essential power is already the normal power source. The GL 90-06 model TS recommendation for cycling the PORV during Mode 3 or 4 is already performed in accordance with an existing DBNPS procedure. The remaining GL 90-06 model TS Action recommendations are not appropriate to the DBNPS multiple vent path design which has its own plant-specific TS requirements in place.
Furthermore, the DBNPS has additional TS Surveillance j Requirements regarding vent path manual isolation valve )
position verification, vent path valve cycling from the control room, and vent path flow verification which go beyond-the requirements proposed in GL 90-06. In conclusion, the present DBNPS TS requirements ensure that the plant is operated safely without undue risk to the health and safety of the public.
A table of the TS requirements for the PORV and block valve is provided in Attachment 2 of this letter. Attachment 2 provides a comparison of the PORV and block valve TS as specified by the model TS of GL 90-06 and the DBNPS TS.
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B. NRC REGULATORY' ANALYSIS AND MODEL TECHNICAL-SPECIFICATIONS :
COMPARISON TO DBNPS DESIGN BASIS
- The model TS for B&V plants provided in GL 90-06 (Attachment A-4) and the supporting regulatory analysis !
provided in NUREG-1316 have also been reviewed by Toledo !
Edison against the DBNPS design basis as established in the DBNPS Updated Safety Analysis Report (USAR). Generic Letter 90-06 and NUREG-1316 (sections 3.2 and 5.3) discuss that the ,
i proposed improvements in GL 90-06 show only a small potential decrease in core melt probability considering reliance on the i PORV for the following design basis safety related functions: l
- 1. Mitigation of a steam generator tube rupture (SGTR) >
accident,
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P
- 2. Low-temperature overpressure protection (LTOP) of the j reactor vessel during startup and shutdown, or j
- 3. Plant cooldown in compliance with BTP RSB 5-1.
Of these three topics, LTOP is not an issue for DBNPS as l noted in GL 90-06 and NUREG-1316. Furthermore, it is important to note that the PORV and block valve are not the l primary means for addressing a SGTR or performing a plant i natural circulation cooldown. While Toledo Edison believes l that the use of other equipment than.the PORV to manage SGTR and plant cooldown functions is an adequate reason for not J adopting the NRC's model TS, each of the above uses of the '!
PORV are discussed for the DBNPS as follows:
I i .
i 1. Mitigation of a Design-Basis Steam Generator Tube Rupture (SGTR) Accident i W In the event of a SGTR accident, RCS leakage into the secondary system may eventually lift the Main Steam Safety j Valves (MSSVs), allowing a release directly to the :
environment. To prevent this situation from occurring, RCS pressure must be decreased to minimize primary-to-secondary j leakage. Minimizing the leak rate vill minimize the SG fill -
rate and the gaseous release rate to the atmosphere. In this ,
event, it is necessary to depressurize the RCS below the ;
i lovest MSSV setpoint so that if the SG fills and becomes ;
, pressurized, the MSSV vill not lift. !
In accordance with emergency procedure DB-0P-02000 (RPS, ;
SFAS, SFRCS Trip or Steam Generator Tube Rupture), control of l the event is accomplished by steaming the SGs through the turbine bypass valves (TBVs) or through the Atmospheric Vent I
Valves (AVVs) if the condenser is not available. The TBVs and AVVs are controlled to assure that SG pressere is kept '
l below the MSSV lift setpoint. The RCS is depressurized by j turning off the pressurizer heaters and using pressurizer j spray if reactor coolant pumps (RCPs) are available. :
t l
Docket Number 50-346 License Number NPF-3 Serial Number 2128
. Attachment 1 Page 11 Forced circulation cooldown is preferable to natural circulation, especially during SGTRs where an expeditious cooldown is desired. Forced circulation eliminates stagnant hot spots that can potentially occur during natural circulation, and thus avoids the complications and cooldown delays that result from void formation in the RCS. In addition, forced circulation provides pressurizer spray, which in turn optimizes RCS pressure control. RCP operation also results in a lower RCS differential temperature which allows a lower primary-to-secondary differential pressure, thus minimizing the tube leak flow rate.
Pressurizer spray can be credited since the analysis does not assume loss-of-offsite power in USAR Section 15.4.2 (Steam Generator Tube Rupture). As a result, the RCPs are available for pressurizer spray. However, if the pressurizer spray is not available, then the use of the pressurizer vent line (valves RC239A and RC200) is directed as the primary means to aid in RCS pressure reduction. The use of the pressurizer vent line (RC239A and RC200) for recovery from a SGTR was chosen because a restricting orifice in this vent line limits flow to provide a more controlled pressure reduction, with less risk of unintentionally rupturing the pressurizer quench tank rupture disk. Since the PORV and block valve flovpath (RC2A and RCll) and the pressurizer vent line (RC239A and l RC200) both discharge to the pressurizer quench tank, the l pressurizer vent line (RC239A and RC200) is the preferred method of depressurization because it minimizes the potential for rupturing the rupture disk of the quench tank which collects the discharge, thereby limiting the potential for contaminating containment. The PORV and block valve flov-path is intended to be only used should the pressurizer spray and pressurizer vent line not be available or if the depressurization rate through the pressurizer vent line is too slov.
The USAR Section 15.4.2 (Steam Generator Tube Rupture) analysis is based on a double-ended tube rupture which results in a RCS leak rate of 435 gpm in the affected steam generator. For this accident scenario, radiation releases were assumed to be terminated from the affected steam generator when RCS pressure falls below the Main Steam Safety Valve setpoint. The PORV vas not credited for accomplishing the P.CS depressurization, rather the RCS pressure drops due to the tube rupture in the steam generator steaming directly to the environment. Steam relief directly to atmosphere continues until the RCS depressurizes below the lovest MSSV setpoint and the affected steam generator is isolated. The RCS cooldown and depressurization continues by further l opening the turbine bypass valves on the unaffected steam generator. During the venting time of the affected steam generator, the radiological consequences were calculated assuming all fission products from the RCS vhich enter the affected steam generator are directly emitted to the
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Attachment 1 Page 12 ;
environment. The doses for the event were calculated with no ,
credit of partitioning of the nuclides in the steam generator (i.e., assuming all isotopic activities is released to the affected steam generator and subsequently released directly ;
to the environment). The calculated doses are well within the 10 CFR Part 100 guidelines with no credit taken for the usage of the PORV for accomplishing RCS depressurization. j The USAR Section 15.4.2 (Steam Generator Tube Rupture)-
analysis also does not specify all of the methods of depressurization. However, the emergency procedure (DB-0P-02000) identifies several such diverse methods to depressurize the RCS for a SGTR:
removal and for keeping SG pressure less than 1000 psig j as the primary method to prevent lifting the MSSVs, ;
- 2) Using normal pressurizer spray, i
- 3) Using the pressurizer vent line (RC239A and RC200) to aid j RCS depressurization, and
- 4) Using PORV and block vent path as a backup to the i pressurizer vent line (RC239A and RC200). ;
The PORV is not the primary method for depressurization for a {
SGTR. Plant procedures give first priority to use of other !
equipment and only resorts to the use of the FORV if other equipment is unavailable or if the depressurization rate '
through the pressurizer vent line is too slow because of non-SGTR complicating factors. It should also be noted that the pressurizer vent line's valves RC239A and RC200 are :
nuclear safety-related "0" valves, tested under the DBNPS l Motor-Operated Valve Reliability and Improvement Program (GL 89-10) and are included in the second Ten-Year Interval Pump ,
and Valve Testing Program. Since the vent path (PORV and ;
block valve) is a backup to the pressurizer vent line (RC239A 4
and RC200) for SGTRs and both of these vent paths are J alternatives to the preferred methods of steaming through the ~
TBVs and AVVs while using normal pressurizer spray, TS 3.4.3 and 3.4.11 provide an appropriate level of TS requirements l for the PORV, PORV block, and pressurizer vent line valves. s This is further supported by the USAR 15.4.2 calculated doses from the SGTR accident which are well within 10 CFR Part 100 guidelines without taking credit for either the PORV and block valve vent path or the pressurizer vent line for RCS ,
depressurization. l 4
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Docket Number 50-346
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- - Attachment 1 Page 13 i
- 2. Low-Temperature Overpressure Protection of the Reactor Vessel !
During Startup or Shutdown For the case of the lov-temperature overpressure protection (LTOP), GL 90-06 and NUREG-1316 states that B&V plants have been excluded from Generic Issue 94; therefore, LCOs for the !
PORV and block are not needed for LTOP considerations. j
- 3. Plant Cooldown in Compliance with Branch Technical Position ,
RSB 5-1 to SRP 5.4.7, Residual Heat Removal (RHR) System. _l l
NUREG-1316 references Branch Technical Position (BTP) RSB 5-1 i in SRP 5.4.7 vhich states: " Current PVRs should have safety 1 grade systems capable of maintaining the RCS in hot shutdown condition for four hours followed by cooldown to-the cold shutdown condition. In Vestinghouse, B&V and CE plants with-PORVs, depressurization of the RCS is accomplished by using a- (
combination of either RCS fluid contraction caused by the l cooldown and heat losses from the pressurizer to ambient or
- by a safety-related PORV." The following addresses RCS l depressurization at the DBNPS during natural circulation ,
using the pressurizer vent pathways (PORV flovpath and the pressurizer vent line).
The DbMPS received its Operating License in April 1977 and, j therefore, is not committed to Branch Technical Position.RSB ,
5-1 in Standard Review Plan 5.4.7, Residual Heat Removal i J
System. The DBNPS is not required to meet the branch ;
technical position regarding safety grade cooldown. ;
In Section 7.0 (Cooldown on Natural Circulation) of DBNPS ;
procedure DB-0P-06903 (Plant Shutdown and Cooldown), the l procedure directs maintaining the plant in hot standby (Mode i 3), if possible, until the Reactor Coolant Pumps (RCP) are i restored. It is preferable to maintain existing RCS j conditions and just remove decay heat with natural !
. circulation until the RCPs become available. The recommended !
procedure is to restore the plant to as near normal {
conditions as possible unless the plant conditions force a natural circulation cooldown. Normal cooldowns (using forced l circulation of the RCPs) are preferred because they can be ;
better controlled and because faster cooldown rates can be l achieved. !
i Uhen natural circulation cooldowns are used, additional q cooldown considerations arise (e.g., lack of normal !
pressurizer spray control, SG shell cooling, reactor vessel l head cooling, idle reactor coolant loop cooling, longer j cooldown times). For instance, if plant conditions force a natural circulation cooldown, procedure DB-0P-06903 (Plant i Shutdown and Cooldown) restricts the cooldown rate (e.g., j 10"F/ hour) in order to prevent forming a steam space (head ;
void formation) in the reactor vessel head. If pressurizer spray is unavailable (e.g., loss of RCPs), the following ;
i 4
Docket Number 50-346
-- , License Number NPF-3 Serial' Number 2128 Attachment 1 i
Page 14 i methods are used for primary depressurization during natural '
circulation cooldowns: (1) manual control of the Turbine Bypass Valves or the Atmospheric Vent Valves to establish the cooldown rate, (2) manual control of the pressurizer heaters. !
to aid in RCS pressure control, and-(3) pressurizer venting through the pressurizer vent line (RC239A and RC200), as necessary. ,
When in the natural circulation mode,-RCS pressure decreases I
(without pressurizer spray flow) result from pressurizer ambient losses and cooldown due to insurges of cooler reactor coolant. As the RCS fluid contracts during the cooldown, the temperature inside the pressurizer is reduced and, thereby, reduces pressurizer pressure. The RCS cooldown rete is controlled using the TBVs or AVVs to prevent the formation of ,
voids in the Reactor Vessel head and to maintain the RCS pressure and temperature within allovable values. Manual- !
control of the pressurizer heaters and pressurizer venting .
through the pressurizer vent line (RC 239A and RC 200) are also used as necessary to aid in RCS pressure control.
The use of the PORV flovpath (RC2A and RC11) for pressure control during a natural circulation cooldown (e.g.,
loss-of-offsite power and loss of RCPs) is not desired at the DBNPS and is not specified in the procedure DB-0P-06903. '
Section 7.0 (Cooldown on Natural Circulation) of procedure-DB-0P-06903 directs the use of the pressurizer vent line (RC239A and RC200). The use of the pressurizer vent line for '
depressurization during natural circulation cooldowns is desired since the pressurizer vent line's depressurization rates through the flovpath of valves RC239A and RC200 are less than those for PORV flovpath venting and, therefore, r more controllable. The pressurizer vent line (RC239A and .
RC200) has a restricting orifice in this vent line to limit ,
flow. Therefore, the pressurizer vent line (RC239A and- ;
RC200) provides a more controllable pressure reduction with less risk of unintentionally rupturing the pressurizer quench +
tank rupture disk. :
Furthermore, using the PORV flovpath (RC2A and RCll) or the I pressurizer vent line (RC239A and RC200) has the characteristic of removing coolant from the system, which.
must be made up (in addition to the contraction experienced during normal cooldowns). Therefore, the pressurizer vent ,
line (RC239A and RC200) is also preferred over the PORV i venting in order to limit the Reactor Coolant inventory.13 be -
made up. As is.the case for a SGTR, the PORV is not the primary method for depressurization for a natural circulation ,
cooldown, and procedures give first priority to the use of i other equipment. It should also be noted that the !
pressurizer vent valves RC239A and RC200 are nuclear .
safety-related "0" valves, tested under the DBNPS '
Motor-Operated Valve Reliability and Improvement Program (GL 89-10), and are included in the second Ten-Year Interval Pump
Docket Number 50-346
. License Number NPF '
Serial Number 2128 Attachment 1 Page 15 and Valve Testing Program. As a result, no changes to the DBNPS PORV and block valve TS operability requirements based on natural circulation cooldown are appropriate.
C. BEYOND-DESIGN BASIS FEED AND BLEED COOLING EVENT As discussed in NUREG-1316, most of the safety enhancement for the proposed backfit is derived from the increase in feed and bleed capability. Section 5.3 of NUREG-1316 specifically recognizes that, without consideration of the function for feed and bleed cooling capability, the recommendations of-GL 90-06 for improving the PORV and block valve reliability are not justified by the regulatory analysis. For the DBNPS, feed and bleed cooling would only be required in a beyond-design basis event involving the loss of primary-to-secondary heat transfer (e.g., loss of both main feedvater pumps, loss of both turbine driven auxiliary feedvater pumps, and loss of the motor driven feedvater pump). The previously existing feed and bleed cooling capability required the use of the PORV, RCS pressurizer code safety valves (PSVs), and both makeup pumps. However, upgrades to the Makeup (MU) System, as described in' Toledo ,
Edison's letter to the NRC dated September 18, 1990 (Serial Number 1836), have provided increased flow capability, train independence, reduction of common mode failure probability, and functionality following a seismic event.and a loss of offsite power. As a result of these upgrades, feed and bleed 1 cooling vill not be lost upon a failure of either a MU pump or the PORV. Analytical results indicate that successful feed and bleed cooling vill be attained with the following minimum equipment combinations:
Two makeup pumps and the RCS pressurizer code safety valves, or One makeup pump operating in piggyback with a Low Pressure Injection Pump, RCS pressurizer code safety valves, and the PORV.
For the DBNPS, these upgrades, therefore, have allowed for the loss of the PORV flovpath without the loss of feed and bleed cooling capability. The loss of the PORV flovpath does not affect the capability to properly cool the core as long as two makeup pumps and the PSVs are available.
The emergency procedure (DB-OP-02000) directs the operation of the MU pumps in the MU/LPI piggyback mode (with the PORV open) with two operating MU pumps even though analysis has determined that the core could be adequately cooled without doing so (i.e., piggyback operation with the PORV open is only necessary for adequate core cooling when only one MU !
pump is operational). The piggyback mode with the PORV open l is stipulated for conservatism since the additional flow to the RCS vill increase the MU/HPI cooling capability and vill I
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provide simplicity of operator actions (i.e., less likelihood for operator error since the operator vill perform the same actions whether one or two pumps are operational).
Additional margin is also available for equipment failures (i.e., with two MU pumps operating, the availability of the i PORV and piggyback could be lost or one MU pump could be lost.
and the core vould still be adequately cooled). Although HU/LPI piggyback with the PORV open and two MU pumps operating is the preferred method for establishing feed and bleed cooling, the flov from two operating makeup pumps 'i through two pressurizer safety valves adequately cools the i core. Vith two makeup pumps operating, failure of the PORV l and piggyback would not impact the capability to adequately '
cool the core by feed and bleed cooling. The DRNPS makeup system has been provided with design upgnaes to prodde feed i and bleed cooling through the pressurizer safety valves with ;
two makeup pumps operating. Accordingly, the PORV is not the j only method for core cooling for events, such as a loss of ;
all feedvater, that are beyond the design basis. Technical ,
Specification operability requirements exist for the two MU pumps (TS 3.1.2.4) and the pressurizer code safety valves (TS 3.4.3). As a result, no changes-to the DBNPS PORV and block l valve TS operability requirements based on feed and bleed cooling are appropriate.
D. COMPARISON VITH NRC INTERIM POLICY STATEMENT ON TECHNICAL .
SPECIFICATION IMPROVEMENTS-The criteria of the Interim Policy Statement on TS !
Improvements (52FR3788) were also evaluated by Toledo Edison [
and the BWOG. The Policy Statement delineates three criteria l vhich establish which constraints on design and operation of l nuclear power plants are appropriate to be in the TS in accordance with 10CFR50.36:
Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant l abnormal degradation of the reactor coolant pressure i boundary. !
Toledo Edison Evaluation of Criterion 1: The PORV and block '
valve are not used to detect a significant abnormal degradation of the reactor coolant pressure boundary.
.l Criterion 2: A process variable that is an initial condition j of a DBA or transient analysis that either assumes-the j failure of or presents a challenge to the integrity of a ;
fission product barrier. ,
Toledo Edison Evaluation of Criterion 2: The PORV and block valve are not process variables.
l
\
l i
Docket Number 50-346
. License Number NPF-3 Serial Number 2128 I Attachment 1 Page 17 Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
Toledo Edison Evaluation of Criterion 3: The design basis accident or transient applicable to Criterion 3 is the SGTR.
I Davis-Besse USAR Section 15.4.2 (Steam Generator Tube Rupture) evaluates this design basis accident. The primary success path to mitigate the SGTR includes depressurization-of the RCS to terminate the leakage of reactor coolant to the l
secondary system. Although the USAR analysis does not explicitly identify all methods used for depressurization, it does assume normal pressurizer spray is available. As described in the emergency procedure, the following systems or components are methods to accomplish this function: (1)-
steaming through the TBVs or AVVs to maintain SG pressure as the primary method to prevent lifting the MSSVs, (2) normal pressurizer spray, (3) use of the pressurizer vent line (RC239A and RC200) to aid RCS depressurization, and (4) use of the PORV flovpath as a backup to the pressurizer vent line (RC239A and RC200). Thus, since several options are available to perform this function, it is inappropriate to identify the PORV and block valve as the primary success path for mitigation of a SGTR. Furthermore, no credit is taken in the DBNPS USAR for the usage of the PORV and block valve vent path or the pressurizer vent line for accomplishing RCS depressurization during a SGTR, and the calculated doses are vell vithin the 10 CFR Part 100 guidelines.
i Risk Significance Provisions: In addition to the above three criteria, the Interim Policy Statement also recommends that i constraints of prime importance in limiting the likelihood or severity of the accident sequences that are found to dominate j risk be included within the TS.
Toledo Edison Evaluation of Risk Significance Provisions: As described above, the regulatory analysis provided in NUREG-1316 relies on use of the PORV and block valve to establish feed and bleed cooling to justify the recommendation of GL 90-06. Since the DBNPS has two steam driven main feedvater pumps, two steam driven auxiliary feedvater pumps, and an electric driven feedvater pump capable of ensuring that water can be delivered to the secondary side of the SGs, the need for feed and bleed cooling is a relatively rare and unlikely event. If the PORV and block valve flovpath is unavailable, the makeup system has the capability to pump against the pressurizer safety valves to provide adequate feed and bleed cooling. Due to the capability of the makeup system to pump against the pressurizer safety valves, use of the PORV for feed and bleed cooling is not a constraint of prime importance in limiting the likelihood or severity of the accident sequences that are
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- Attachment 1 Page 18 j found to dominate risk. Furthermore, the DBNPS Probabilistic- l Risk Assessment indicates the effect of the PORV flovpath on !
t cooldown (SGTR, plant cooldown by natural circulation, and feed and bleed cooling) to be so small as to be considered l insignificant.
l It is concluded from the above evaluation of the criteria in the NRC Interim Policy Statement and in GL 90-06 for safety l functions that changes to the DBNPS TS are inappropriate.
E. ADDITIONAL INFORMATION REGARDING GL 90-06 MODEL TECHNICAL -f SPECIFICATIONS i While Toledo Edison believes that the use.of other equipment ;
than the PORV to mitigate a SGTR, provide for a plant !
cooldown, and perform feed and bleed cooling is an adequate .
justification for not adopting the NRC's GL 90-06 model TS, !
the following information is also provided. I
- 1. The one or two hour restoration time for an inoperable ,
PORV or block valve is disproportionately short relative !
to the frequency of the challenge (i.e., the frequency of i the need for the PORV for SGTR, plant cooldown, or for ,
feed-and-bleed). Since the PORV is not the first choice ~ l 4 for a SGTR or plant cooldown, LTOP is excluded in ;
GL 90-06 for B&V-designed plants such as the DBNPS, and j the flow from the two makeup pumps through the two- .t pressurizer code safety valves provides an alternative l means for' feed and bleed cooling capability, Toledo i Edison maintains that a change to the present DBNPS TS j 1 3.4.3 and 3.4.11 restoration / shutdown for the PORV is ;
inappropriate.
- 2. The one or two hour restoration / shutdown may not provide !
adequate time to perform an operability determination- l (e.g., degraded, but operable) in accordance with GL !
91-18 (Information to Licensees Regarding Two NRC' ;
Inspection Manual. Sections on Resolution of De raded and i Nonconforming Conditions on Operability). To impose one :
and two hour restoration / shutdown requirements and !
subject the overall plant to a cooldown and heatup .
transient is unvarranted for an event eyond the design ;
basis, particularly when an alternate feed and bleed ,
method (i.e., two makeup pumps pumping against the ;
pressurizer code safety valves) is available. -l l
- 3. The use of a one or two hour restoration / shutdown time '!
vould lik.ely result in the need for Toledo Edison to request enforcement discretion from the NRC staff on an ;
expeditious basis. With the NRC. staff's guidance to i first seek enforcement discretion, if appropriate, the !
one or two hour shutdown requirement should be regarded j as impractical. Based on the DBNPS's RCS multiple i l
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venting design it vould be appropriate to request .
enforcement discretion in the case of~an inoperable block -;
valve or PORV under the requirements of the GL 90-06 !
model TS. ;
- 4. The calculated cost of the proposed shutdown requirement ,
appears to be underestimated in NUREG-1316 Section 5.4, !
Outage Avoidance Cost, and in Section 5.5, Cost / Benefit Comparison. There is a high potential for substantial .
costs in replacement power due to outages resulting from the' proposed one-hour.and two-hour shutdown requirements. !
If failure of the PORV or the block valve vould occur !'
during power operation under the proposed model TS, DBNPS would be required to expeditiously shutdown. EIn order to- i restore the valves to operability, the unit vould likely !
have to be placed in Cold Shutdown (Mode 5) due to the l environmental conditions in the general area of the -i valves. ,
NUREG-1316, Section 5.4, Outage Avoidance Cost, refers-to three Electric Power Research Institute (EPRI) reports. The first two EPRI reports are referred to in calculating the average plant capacity losses due to PORV ;
and block valve problems. Section 5.4 references these 3 reports stating " ... PVR capacity losses, that is, ;
outage time due to RCS relief valve problems'have i remained relatively unchanged over the years," and then ,
uses the report's RCS safety / relief valve caused plant !
capacity loss values as the basis in calculating the Outage Avoidance Cost as a benefit in implementing'the ,
recommendations of Generic Letter 90-06. Section 5.4 further states that the EPRI reports " ... infer (s) that little, if any, improvement has been made in the maintenance and quality assurance procedures for these valves". Since the EPRI data from these two reports ,
(published 1981 and 1984, respectively) were for the 3 years 1968 - 1979 and 1980 - 1982, the data may.not be !
relevant to today's plant operating experience, and maintenance and quality assurance programs.
The third EPRI report referenced in Section 5.4 is also used to further the recommendations of Generic Letter r 90-06 (including additional TS on the PORV and block -;
valve) and to reach a conclusion that an approximate 75 ,
percent reduction in plant capacity losses due to PORV and block valve problems could result by incorporating- '
the Generic Letter recommendations. This EPRI report,
" Limiting Factor Analysis of High Availability Nuclear Plants, Volume 3: Supplemental Report, Limiting Valves Study," EPRI NP 1139, was published in 1979 and its applicability to the performance of plants today may also not be relevant.
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Docket Number 50-346
. License Number NPF-3 -1 Serial Number 2128 l
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Section 5.4 uses the EPRI data to calculate' a plant ,
capacity loss per year. This capacity loss factor is- !
then'used along with the 1979 EPRI report's reduction in capacity losses to calculate the Outage Avoidance Cost ;
(the cost savings from avoiding an_ outage due to poor i PORV or block valve performance). The Outage' Avoidance ,
Cost formula is highly sensitive to the values _used in estimating the capacity losses and the replacement power -i costs. For example, NUREG-1316 used a replacement-power :
cost estimate of $500,000 per day. However, the l replacement power costs for the DBNPS given in i NUREG/CR-4012 (Replacement Energy Costs for Nuclear '
Electricity - Generating Units in the United States: ;
1992 - 1996) for Vinter 1991 - 1992 is $146,000 per day .i (end-of-year 1991 dollars). Using this value gives an :
Outage Avoidance Cost of $48,000 per reactor-year instead 1 of NUREG-1316's cost of $165,000 per' reactor. year, i Present value with a discount rate of 5 percent for a 1 period of 25 years (remaining life on the DBNPS 40-year ,
license) is approximately $660,000 as compared to. i NUREG-1316's present value of $2,541,000. Accordingly,. ;
the cost of a single forced outage of four days due to l the proposed model TS requirements for an inoperable PORV 6 or block valve during the remaining DBNPS plant life 1 vould completely exceed the calculated,0utage AvoidanceL l Cost. This credible scenario of a required plant i shutdown due to the proposed model TS is a negative impact and is not addressed in NUREG-1316 and the associated cost justification for this backfit.
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F. OVERALL CONCLUSION Toledo Edison is in general agreement with NRC Recommendation Items 3.1.1 and 3.1.2 provided in GL 90-06. The DBNPS ,
Quality Assurance Program, Pump and Valve Testing Program (in i conjunction with procedure DB-SP-03363), and the maintenance i and maintenance training programs will ensure' reliable PORV j and block valve operation. Regarding NRC Recommendation !
Item 3.1.3, the existing DBNPS TS for the PORV and block ;
valve contain the appropriate shutdown requirements for the DBNPS system design and operation.
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noctet Number 50-346 License Number NPP -3
, Serdal Number 2128 Attachment 2 Page 1 COMPARISONS OF PORV AND BLOCK VALVE TECHNICAL SPECIPICATIONS GL 90-06 DBNPS GL 90-06 MODEL TS DBNPS TS RESPONSE Limiting Condition for TS 3/4.4.11 (RCS-RCS Retain-present TS, but Operation for the PORV Vents) expand Bases. A and its associated Block Technical Specification Valve change to Bases 3/4.4.ll was submitted to the NRC Applicability: Modes 1, Applicability: Modes 1, by Toledo Edison Serial 2, 3 2, 3 Letter 2046 to add more TS 3.4.4 Action a: Por Action for an inoperable discussion of the excessive PORV seat pressurizer vent path functions of the PORV and Icakage, restore the PORV through both " valves RC11 block valve in the DBNPS -
to operable status or and RC2A (PORV)" and system configuration.
close the block valve " valves RC239A and RC200 (with electrical power (pressurizer vent line)," With the inoperable PORV maintained on) within one and with the vent path in flovpath (RC11 and RC2A),
hour. Or otherwise be in each RCS Loop operable: a second pressurizer vent Mode 3 in the next six Restore the pressurizer path (pressurizer vent hours and Mode 4 in the vent path through either line through RC239A and following six hours. valves RC11 and RC2A RC200) remains available (PORV) or valves RC239A for the steam generator and RC200 (pressurizer tube rupture event and vent line) within 30 days plant natural circulation or be in Mode 3 in six cooldowns. The use of hours and Mode 4 in the the two pressurizer Code following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. safety valves with two makeup pumps operating is The DBNPS TSs do not adequate for promoting require the removal of feed and bleed cooling electrical power from a without the PORV and block valve closed due to block valve path being excessive PORV leakage. available. There are TS requirements for the makeup pumps in TS 3.1.2.4 and the Code safety valves in TS 3.4.3.
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Ncket Number 50-346
+ Licqnse Number NPF -3
, Serial Number 2128 Attachment 2 Page 2 COMPARISONS OF PORV AND BLOCK VALVE TECllNICAL SPECIFICATIONS GL 90-06 DBNPS GL 90-06 MODEL TS DBNPS TS RESPONSE TS 3.4.4 Action b: For Existing DBNPS TS 3.4.3 Same as above the PORV being (RCS - Safety Valves and inoperable, for reasons PORV - Operating) other than excessive seat requires that if the PORV leakage, restore the PORV becomes inoperable it to operable status or either is isolated or TS close the block valve 3.0.3 is entered with within one hour with its action initiated within i electrical power removed. one-hour to shutdown the If the block valve is plant, be in at least closed, be in Mode 3 in Mode 3 within the next ,
the next six hours and six hours, and be in at Mode 4 in the following least Mode 4 within the six hours. following six hours. As described above, the Action Statement #
requirements of TS 3/4.4.11 also apply for the multiple RCS vent paths.
TS 3.4.4 Action c: If the Same as for TS 3/4.4.11 Same as above -
block valve is requirements previously inoperable, within one described above.
hour restore the block valve to operable status or the PORV must be placed in manual control and the block valve restored in the next hour, or the plant placed in Mode 3 vithin the next six hours and in Mode 4 within the following six hours.
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Qooket Number 50-346
- Licqnse Number NPF -3
, Ser-lal Number 2128
-At tachment 2 Page 3 COMPARISONS OF PORV AND BLOCK VALVE TECHNICAL SPECIFICATIONS GL 90-06 DBNPS GL 90-06 MODEL TS DBNPS TS RESPONSE TS 3.4.4 Action d: 3.0.4 3.0.4 NA for TS 3/4.4.11 The GL 90-06 model TS WA (Can startup plant by (Can startup plant by would allow plant startup relying on an Action relying on an Action while relying on an Statement) Statement). However, Action statement (i.e.,
3.0.4 applies (cannot TS 3.0.4 is not ,
startup plant relying on applicable). However, an Action statement) for DBNPS TS 3.0.4 is TS 3/4.1.2.4 (Makeup applicable to TS Pumps) and TS 3/4.4.3 3/4.1.2.4 regarding the-(Code Safety Valves and Makeup pumps and to TS PORV). 3/4.4.3 regarding the pressurizer code safety valves. i Surveillance Requirements ,
(SRs)
SR 4.4.4.la: Operate the TS 4.4.11.2: Cycle the The testing requirements ,
PORV through one complete PORV through one complete of the PORV to meet TS cycle in Modes 3 or 4 at cycle in Modes 5 or 6 at 4.4.11 and TS 4.0.5, in- .
least once per 18 months, least once per 18 months. conjunction with the in addition to the procedurally - required requirements of cycle test (procedure Specification 4.0.5. DB-SP-03363) of the PORV in Mode 3, represents the PORV's ability to reliably perform its function.
SR 4.4.4.lb: Perform TS 3/4.4.3 (Safety Valves DBNPS TS 4.4.3 fulfills Channel Calibration of and PORV Operating): For the requirements of SR actuation instrumentation the PORV, perform a 4.4.4.lb of the model TS at least once per 18 Channel Calibration check of GL 90-06.
months. every 18 months.
SR 4.4.4.2: Cycle the Cycle the block valve Model TS for Block Valve block valve at least once thru one complete cycle testing (SR 4.4.4.2)-is per 92 days to in accordance with TS the same as DBNPS TS demonstrate the valve's 4.0.5 and the Inservice 4.0.5 testing-operability. Testing Program every 92 requirements.
days.
Ddeket Number 50-346
' Licelise Number NPF -3 l* Serial Number 2128 Attachment 2' Page 4 COMPARISONS OP PORV AND BLOCK VALVE TECHNICAL SPECIFICATIONS GL 90-06' DBNPS GL 90-06 MODEL TS DBNPS TS RESPONSE-SR 4.4.4.3: Demonstrate None This model TS.
surveillance requirement PORV and block valve emergency power supply is not appropriate to the operability by DBNPS design since Class
. transferring motive and lE essential power is control power from normal already is the normal.
to emergency power and power supply.- This is-operating the valves recognized by the_GL' through a complete cycle 90-06 model TS Bases of full travel at least applicable to once per 18 months. B&V-designed plants such' ;
as DBNPS (Attachment A-5).
Verifying all manual DBNPS TS requirements isolation valves in each beyond the GL 90-06 model RCS vent path are locked TS.
in the open position at least once per 18 months I (TS 4.4.11.1).
Cycling each valve in the DBNPS TS requirements RCS vent path through at beyond the GL 90-06 model~
least one complete cycle TS.
of full travel from the control room during Mode 5 or Mode 6 at least once per 18 months (TS 4.4.11.2). .
Verifying flov through DBNPS TS requirements ,
the reactor coolant vent beyond the GL 90-06 model system vent paths during TS.
Mode 5 or Mode 6 at least once per 18 months (TS 4.4.11.3).
TS 3/4.1.2.4 does not DBNPS TS requirements allow plant startup if a beyond the GL 90-06 model' Makeup Pump is TS.
inoperable. -i
TS 3/4.4.3 does not allow DBNPS TS requirements plant startup if a beyond the GL 90-06'model ,
pressurizer Code safety TS.
valve is inoperable.
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