ML20128F117
| ML20128F117 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/29/1993 |
| From: | Broughton T GENERAL PUBLIC UTILITIES CORP. |
| 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 C311-93-2003, GL-90-06, GL-90-6, NUDOCS 9302110249 | |
| Download: ML20128F117 (5) | |
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OPU Nuclear Corporation NQQIgf Post Office Dox 480 Houte 441 South Middletown, Pennsylvania 17057 0191 717 944 7621 TELEX 84 2386 Writer's Direct Nal Number-717-948-8005 1
January 29, 1993 C311-93-2003 i
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1 U. S. Nuclear Regulatory Commission Attention:
Document Control Desk Washington, D.C.
20555 Gentlemen:
Subject:
Three Mile Island Nuclear Station Lait 1 (THI-1)
Docket No. 50-289 Operating License No. DPR-50 GPU Nuclear Response to NRC Request for Additional Information Regarding Generic Letter (GL) 90-06, " Resolution of Generic Issue 70, ' Power-0perated Relief Valve and Block Valve Reliability,'and-Generic Issue 94, ' Additional Low-Temperature Overpressure Protection for Light-Water Reactors.' pursuant to 10 CFR 50.54(f)."
GPU Nuclear responded to Generic Letter (GL) 90-06 in a letter dated December 24, 1990. The stated purpose of GL 90-06 was to provide technical resolution of two generic issues and impose plant backfits which were considered to be cost-justified safety enhancements. Generic Issue 94 did not apply to THI-1.
In' regard to Generic Issue.70, " Power-0perated Relief Valve and Block Valve Reliability," the GPU Nuclear response _ addressed each item as.
required by the generic letter.
The NRC. staff reviewed.the GPU Nuclear respor.se and by letter dated October 29, 1992.provided comments regarding the incorporation of model Technical Specifications and testing of the-Power-0perated Relief Valve (PORV).
NRC Request:
.The staff will not accept, without sufficient justification,.the position.that the technical specification upgrades.are unnecessary because.the PORVs are not the primary n..;ans of dealing with the three safety functions identified in the generic letter.
GPU Nuclear Response:
Because the application of-model Technical Specifications. is-a generic :issu)
- that affects all other B&W operating plants, the B&W Owners Group-(BWOG)
Technical Specification Connittee undertook an initiative to respond as-an' k
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C311-93-2003 Page 2 of 4 owners group. A teleconference involving the BWOG utilities and the NRC staff I
was held on December 17, 1992. As a result of that discussion, a BWOG res)onse, addressing the model Technical Specifications, has been prepared wit i THI-l participation. The BWOG letter (0G-1128), dated January 18, 1993 is attached.
NRC Request:
1he staff position requires the 18-month PORV stroke test to 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.
Your submittal did not adecuately meet this staff sosition.
The staff is not accepting Mode 5 (COLD St VTDOWN) testing siinply
>ecause it is allowable by the ASME Code or that the NRC approved IST program includes Mode 5 for this particular test...
We note that one licensee has proposed the option to bench test the PORVs.
This would be acceptable, provided the tests are performed at conditions simulating Mode 3 or 4 conditions or greater and provided the proper reinstallation of the PORVs and controls is verified...
In summary, the staff maintains its position that the-PORVs should be stroke tested during 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.
GPU Nuclear Response:
THI-l does not require the PORV to achieve feed and bleed core cooling.
Because the plant has high capacity, high head, Makeup /High Pressure Injection (HP!) Pumps, feed and biced core cooling can be accomplished at THI-l using-only the pressurizer code safety valves.
The safety function of the TMI-1 PORV is to protect the Reactor Coolant System
.(RCS) from overpressure during low temperature RCS conditions. :Because the-PORV provides a safety function, stroke testing is' required by the THI-l Inservice Testing (IST) Program. Stroke testing of the-TMI-l PORV is currently being performed during refueling at Wyle Laboratories in accordance with GPU Nuclear Specification (SP) 1101-12-087. This test is performed at conditions that are representative of the PORV operating environment. THI-l General Maintenance-Procedure 1401-2.1, " Pressurizer Relief Valve Removal / Installation," provides verification of-~ proper reinstallation of the PORV and controls.
THI-l_ experience has shown-satisfactory results.
Laboratory bench testing each refueling interval _(up to 24 months) is the method preferred by GPU Nuclear for performing a' stroke test of the'PORV to meet ASME Section XI and Technical Specification IST requirements; however, the IST procedure does permit actuation of the PO U in place as an alternative test method.
In place stroke testing of the PORV using steam with RCS temperature >332*F and RCS pressure at 500-600 psig is an acceptable 4
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C311-93-2003 Page 3 of 4 alternative; the test is performed at conditions equivalent to or greater than Mode 3 and Mode 4.'
Stroke testing of the PORV during Cold Shutdown conditions (Mode 5) would not fulfill THI-l IST requirements.
CPU Nuclear prefers laboratory bench testing to testing in-place on the pressurizer for several reasons.
Bench testing is performed at normal steam inlet conditions and verifies both setpoints; 2450 psig for power operation and 485 1sig for low-temperature overpressure protection (LTOP).
Bench testing allows use of a cleaner fluid environment as opposed to RCS fluid i
which may cause deleterious effects on the valve internals and reduce-PORV reliability. The measurement of the PORV stroke time and actual main disc movement can be verified easily under laboratory bench testing conditions whereas stroke time and main disc movement are inferred by indirect indications during in situ testing.
The THI-l PORV is an electromatic, solenoid actuated, pilot operated relief valve which requires a minimum pressure of 50 psig under the main disc for-the PORV to open. As a result, there are several disadvantages to in-place testing of the PORV during hot standby (Mode 3) or hot _ shutdown (Mode 4).
To test the PORV in-place, the upstream PORV Block Valve must be open to supply the necessary fluid pressure (RCS-pressure) through the pressurizer to open the valve. The PORV design does not provide direct stem position indication.
Therefore valve lift must be inferred from alternate indications (tailpipe AP, tailpipe AI, acoustic monitor, RCS pressure decrease,.or Reactor Coolant Drain Tank (quench tank) pressure, temperature, or level rise).
In situ testing of the PORV would also result in an insurge.of cooler water from the hot leg into the pressurizer.
The resulting thermal cycle on the-pressurizer surge line would not be expected to contribute to the' effects of thermal stratification and ther.nal striping of the' surge line as _ described in NRC Bulletin 88-11. However, thermal cycles of the surge line, even those of-low magnitude, are considered undesirable and should be avoided.
Safety measures have been taken to reduce the challenges to the PORV. The
)ower. operation setpoint of the PORY was raised to 2450 psig; i.e., above the ligh pressure reactor trip setpoint (2355 psig).
TMI-l also has two anticipatory reactor trips: " reactor trip on turbine:trir with reactor power
-greater than 45% power" and " reactor trip on loss of bot 1 feedwater pumps with reactor power greater than 7% power."
It was a consideration in choosing setpoints for the anticipatory trips to avoid actuation of the-PORV.- These measures have been successful in reducing challenges to the PORV.- ' Requiring that the PORV be tested in-place at Hot Standby or llot Shutdown conditions would ' unnecessarily: increase-the number of challenges to the PORV. On this basis, in-place testing of the PORV would result -in an overall reduction in
- NUREG-1430,." Standard Technical Specification _ for ' Babcock : andJilcox Plants," Rev 0. September 1992.. THI-l has custom technical specifications, where.
plant operating conditions do not correspond to the numerical modes definitions of the Standard Technical Specifications (STS).
d 011-93-2003 Page 4 of 4 plant safety.
Unless there is a need to reestablish the operability of the PORV, performing an in-place PORV test would nnt be necessary because the valve will have satisfactorily completed a bench test prior to its installation during an outage.
Performing a bench test under controlled conditions-allows the opportunity for accomplishing repairs to the valve while there is sufficient time for any needed repairs or additional testing without these activities affecting the unit's operating schedule and without the additional-dose that
- could result unnecessarily by having to remove or perform work on the valvo while it is attached to the RCS.
If a test failure during plant startup were to require cold shutdown conditions, this would result in an additional thermal cycle on the unit and additional personnel exposure that could otherwise be avoided.
For these reasons it is not desirable from a plant safety perspective to perform in-place testing of the PORV instead of or in addition to the current method of bench testing.
In-place testing also increases the likelihood that the valve will develop seat leakage during the operating cycle.
GPU Nuclear has concluded that bench testing the PORV, as described above, is the preferred test method, llowever, if it becomes desirable to take credit for an in-plant test, the current
)rocedures permit that option.
In summary, GPU Nuclear cor.cludes that althougi the PORV does not have a safety related function during nnwer operation, the current THI-1 procedures for stroke testing of tM. FORV provide adequate assurance of PORV reliability for operating r, well as shutdown conditions.
Sincerely, hh>
T. G. Broug on Vice President and Director, THI-l MRK Attachment cc: Administrator, Region I THI-1 Senior Resident Inspector THI-l Senior Project Manager s
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C311-93-2003 METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGilT COMPANY PENNSYLVANIA ELECTRIC COMPANY GENERAL PilBLIC UTIlllIES NUCLEAR CORPORATION Three Mile Island Nuclear Station,' Unit 1 (THI-1)
Operating License No. DPR-50 Docket No. 50-289 GPU Nuclear Response to NRC Request for Additional Information Regarding Generic Letter 90-06 This letter is submitted in response to the NRC letter of October 29, 1992 which requested additional information regarding Generic letter 90-06. All' statements contained in this response have been reviewed, and all such statements made and matter set forth therein are true and correct to.the best of my knowledge.
b T.G.BroughtQ Vice President and Director, THI-l i
Sfgned and sworn before me this M -day of-
'W wu1AL M, 1993.
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GPU Nuclear Csvparation TMI-t B& W Nuclear Technategins Working Together to Economically Provide Rehable and Safe Electncal Power Suite 525 e 1700 Rockville Pike e Rockville. MD 20852 e i301) 230-2'00 January 18,1993 0G 1128 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555
Subject:
Davis Besse, Docket No. 50 346 Crystal River 3, Docket No. 50-302 Oconee Nuclear Station, Docket Nos. 50 269,270,287 Arkansas Nuclear I, Docket No. 50-313 Three Mile Island Unit 1, Docket N. 50-289 B&WOG responte to " Staff Revievi of Generic Letter 90-06," Resolution of Generic issue 70,' Power-Operated Relief Valve and Block Valve Reliability,'
and Generic Issue 94,' Additional lxw Temperature Overpressure Protection -
for Light Water Reactors,' Pursuant to 10CFR50.54(f)"", Dated October 29, 1992, signed by Ronald W. Hernan. Note: The referenced letter was sent to GPUN for TMI 1; similar letters have been sent to Davis Besse and ANO 1.
Gentlemen:
BACKGROUND All B&WOG utilities with operating plants responded to Generic Letter 90-06 in December, 1990. The staff has reviewed those responses and provided feedback in correspondence to GPUN (TMI 1), Toledo Edison (Davis Besse) and Entergy (ANO 1). The NRC staff requested this letter in a telecon on December 17,1992. There are two principle issues where tu B& Wee utilities have different positions than the staff: 1) PORV stroke testing; and 2) proposed Technical Specification requirements leading W shutdown for an inoperable PORV and block valve. The stroke testing issue will be addressed by the individual utilities.
The B&WOG is providing a generic response to the staff for the shutdown topic. The issue is:
The B&WOG utilities have declined to make certain technical specification changes as presented in the generic letter. In particular the B&WOG utilities do not believe the shutdown requirements for an inoperable PORV or block valve, shown in
- example Technical Specification 3.4.4, are appropriate measures. The staff states that they "will not accept, without sufficient justification, the position that the technical ation upgrades are unnecessary because the PORVs are 'not the primary g j g g/ q m
0G-1128 January 18,1993 page 2 means of dealing with the three safety functions identified in the generic letter." The three safety functions noted by the generic letter are:
1.
Mitigation of a design basis tube rupture accident.
2.
LTOP 3.
Plant cooldown per RSB 5-1 to SRP 5.4.7 (RHR System).
Of the three functions, LTOP is not an issue for the B&WOG plants as noted in G.L 90 06 ani NUREG-1316. Therefore use of LTOP as justification for imposing a mode change requirement for an inoperable PORV is inappropriate and inconsistent with G.L 90-05.
Although the NRC staff has denied the ANO.1, TMI-1, and Davis Besse positions, we maintain that the use of equipment other than the PORV to manage SGTR and plant cooldown is an appropriate basis for not requiring shutdown for an inoperable PORV.
The staff has further iecognized that "most of the c.afety enhancement for the proposed backfit is derived from the increase in feed and bleed capability." The B&WOG notes that our plants have the ability to provide feed and bleed core cooling through the pressurizer safety valves without use of the PORV. An inopeiable PORV does not degrade tin:
capability to cool the core, and therefore we do not regad this as an adequate reason to require shutdown. We request the staff's reconsideration of their position based on this -
information.
The PORV desi9a at all B&WOG plants is an electromatic D.C solenoid actuated pilot operated relief vre. B&W plants do not use air actuated PORVs. One PORV is installed, on the top of the pressurizer as are two spring loaded safety valves. The PORV piping has an.A.C r.stor operated r.armally open gate valve in.".alled betweer, the PORV and the presmrizer. The PORV and block valve have the capabili:y of being supplied from vital power n required by NUREG 0737, and also have a si nsor at the PORV discharge for detecting an cpm PORV.' PORV reliet capacities vary slightly, but a typical rating is about 105,000 lbm/ hour of steam; some plants are slightly larger. A seperate pressurizer vent line, which has a typical capacity about 1/6 as large as the PORV is also installed on the pressurizer.
B&WOG Position on Technical Soccification Changes The typical present BWOG plant Technical Specification requirement for an inoperable PORV requires block valve closure. Some variance exists among the plants, but no B&WOG plant has Technica1 Specifications requiring shutdown for an inoperable PORV.
The B&WOb maintains that shutdown for an inoperable PORV is not necessary because:
1.
The PORV is not the primary method for depressurization for a SGTR. Plant specific emergency cperating procedures (EOPs) nased on the B&WOG Abnormal.
Transina Operatirg Guidelines (ATOG) concept give first priority to use of other-equipment and er.ty resort to use of the PORV if the r.'her equipment is unavailable.
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~OG-1128 January 18,1993 page 3 To depressurize for a SGTR the preferred method is normal pressurizer spray, driven by reactor coolant pump head rise. In the absence of normal pressurizer spray, some plants have high pressure auxiliary pressurizer spray provided by the makeup system, other plants use the normal pressurizer vent. Either the auxiliary spray.or the pressurizer vent is preferred to the PORV because smoother depressurization control can be afforded as compared to the more abrupt depressurization of the PORV.
The potential for blowing the quench tank (reactor coolant drail tank) rupture disc is greater when the PORV is used. The EOPs also require steaming the steam -
generators to maintain RCS heat removal either for keeping RCS pressure below the secondary relief pressure for the affected generator or to help RCS depressurization; therefore steam generator heat removal can also be used for depressurization.
2.
No B&WOG operating plant has the assumption of loss-of-oifsite power coincident with SGTR as part of the design basis accident licensing analysis presented in the FSAR. As a result the RC Pumps are available for pressurizer spray. The FSAR does not specifically cite the method for depressurization to be used. All plants have provisions for managing SGTR if off site power is not available even though this is not a design basis requirement.
3.
No B&WOG operating plant is committed to the provisions of St,ndard Review Plan 5.4.7 and its Branch Technical Position RSB 5-1 as a part ofits License. We believe it is inappropriate to backfit this requirement via a Generic Letter requesting a PORV Technical Specification.
4.
As noted before, B&WOG plants do not require the PORV for feed and bleed core cooling. B&WOG plants (except for Davis Besse) have very high head /high flow Makeup /HPI pumps that can provide flow to the pressurizer safety valves for feed and bleed core cooling if the PORV is not available. The Davis-Besse _ Makeup system (which is separate from the HPI system) has been redesigned so that it can provide feed and bieed core cooling through the pressurizer safety valves. Various evaluations have demonstrated this design capability.
For example an NRC sponsored study states "one HPI pump delivering flow at the setpoint pressure of the pressurint SRV's is sufficient to prevent core uncovery if initiated by 2400 s." (See NUREG/CR-4966, Volume 2, page 18).
As such the PORV is not the only method for core cooling for events, such as loss of all feedwater, that are beyond the design basis.
5.
The importance of the need for the PORV seems to be exaggerated by the shutdown requirement of the proposed Technical Specification. NUREG 1316 references a Brookhaven National Laboratory (NUREG/CR-4999) study that estimated the risk reduction from improved PORV and block valve reliability. That study showed only -
a small potential decrease of core melt frequency due to increased valve reliability (appro-imately 1 to 3 E-7). The study did not include feed and bleed. Separately, x
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OG-1128 January 18,1993 page 4 I
however, the staff notes that most of the safety enhancement for the proposed backfit is derived from the increase of feed and bleed capability.
The B&WOG contends that the improvement 'n plant safety is insignificant. For example, the CR-3 Probabilistic Risk Assessment (PRA) demonstrates that the n
expected improvement in PORV reliability does not have a significant impact on core damage frequency at CR-3. The reduction in risk due to the estimated improvement in PORV and block valve reliability is minimal. The CR-3 PRA indicates, assuming an optimistic 75% increase in PORV and block valve reliability, that the resulting decrease in core damage frequency is approximately 4.5x10' per year. This is equivalent to a 3% decrease - from approximately 1.5x105 per year to 1.455x105 per year, and is due to the role of the PORV in depressurization during a steam generator tube ruptur.. event. Enhuncement of feed.and-bleed cooling has no effect on core damage frequency for CR-3.
Other B&WOG PRAs indicate similar results:
The Oconee PRA did not expheitly model the PORV for feed arm bleed cooling, since only one of three relief paths (two SRVs, one PORV) are required for success. However. explicit consideration of an inoperable PORV produces a negligible increase of plant risk.
The TMI-1 and ANO 1 PRAs also show insignificant differences of results for an inoperable PORV.
The Davis-Besse PRA indicates the effect of the PORV on cooldown (SGTR, natural circulation cooldown, and feed and bleed) to be so small es to be considered trivial.
6.
Cost evaluations of the propor ad shutdown requirement appear to be underestimated by the NUREG 1316 cost / benefit analysis. There is a high potential for substantial costs m replacement power due to outages resulting from the proposed shutdown requiremem.
For failures that might occur under the propoced Technical Specitication, a shutdown in one hour would be required. If, after shutdown, work on the PORV itself was necessary to restore the valve to operable status, the unit would likely have to be placed in cold shutdown (MODE 5) due to the environmental conditions in the general area of the valve. Such an outage would be expected to typleally last a minimum of 10 days (costing 7-10 million dollars in terms of' r; placement power costs).
7.
The three selection criteria of the NRC's Interim Poliev Statement un Technical Specification Improvement do not appear to have been reviewed by the staff in order to make their decision on what constraints should be put on the PORV in regard to -
Technical Specifications. The B&WOG has reviewed the PORV under the three -
. criteria and find it does not meet any of the three:
4 OG 1128 TJanucry 18,1993 page 5 Criterion 1: installed instrumentation that is used to detect, and indicate in the control toom, a significant abnormal degradation of the reactor coolant pressure boundary.
B&WOG Evaluation of Criterion 1: The PORV is not used to detect a significant abnormal degradation of the reactor coolant pressure boundary.
Criterion 2: A process variable that is an initial condition of a DBA or Transient Analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
B&WOG Evaluation of Criterion 2: The PORV is not a process variable.
. Criterion 3: A structure, system, or component that is part ot' 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.
B&WOG Evaluation of Criterion 3: The DBA or Transient opplicable to.
Criterion I is the SGTR. As noted previously, the safety analysis _ basis _ for SGTR is not explicit for the method of depressurization but normal.
pressurizer spray can be assumed to be available since the FSAR analysis does not assume loss of off site power. In addition, based upon the ATOG concept if normal spray is not available the auxiliary spray or pressurizer vent are preferable to the PORV for this event. Thus, since several_ options are available to perform this function and since use of the PORV during a SGTR is not a-design basis requ_irement it is inappropriate'to identify the PORY ~ s a
the crimary success path for mitigation of a SGTR.
Risk sluificant orovisions: In addition to the three criteria, the Interim Policy.
Statement also recommends that constraints of prime importance in _ limiting the likelihood or severity of the accident sequences that are found to dominate risk be included within the Technical. Specifications.
B&WOG evaluation of risk sienificance provisions: The B&WOG elected to
- include - the PO.RV in the Technical Specifications.because of its. risk
_ The risk element' of interert is prevention' of a small break.
importance.
LOCA via the PORV path caused by a failed open PORV. This is not the..
O same as the elements'(SGTR, feed and bleed, etc.) covered by the staff in G.L 90-06 or NUREG 1316. The B&WOG bcsis is docurnented in the new Standard Technical Speifcations- (NUREG :1430, September.1992).
Prevention of a smat - ' k ~LOCA via the PORV does not require PORV-operability for bpening control.
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OG-1k28 January 18,1993 ~
page 6 Thus the criteria of the Interim Policy Statement do not support the shutd'own-request of the staff. The risk basis for inclusion of the PORV in the Technical Specifications is not the same as that cited by the staff in G.L 90 06.-
For these reasons the B&WOG does not believe it is appropriate to shutdown for an inoperable PORV or block valve as given by the staff in the proposed Technical Specification. If any clarification is needed or more information is required the B&WOG will be pleased to provide it.
Very truly yours, ll l g
Blair Wunderly, Chairman -
B&WOG Technical Specification Committee xc: B&WOG Technical Specification Committee members B&WOG Utility Licensing Managers B&WOG Steering Committee members 4
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