ML20044F984
| ML20044F984 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 05/21/1993 |
| From: | Hagan J Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20044F985 | List: |
| References | |
| NLR-N93054, NUDOCS 9306010285 | |
| Download: ML20044F984 (8) | |
Text
O PSEG Pubhc Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Nuclear Department MAY 211993 NLR-N93054 LCR 92-06 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
LICENSE AMENDMENT APPLICATION SAFETY RELIEF VALVE TESTING REQUIREMENTS HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 This letter submits an application for amendment to Appendix A of Facility Operating License NPF-57 for the Hope Creek Generating Station, and is being filed in accordance with 10CFR50.90.
Pursuant to the requirements of 10CFR50.91(b) (1), a copy of this request for amendment has been sent to the State of New Jersey.
Currently, the Technical Specifications require that at least one half of the Safety / Relief Valves (SRVs) shall be removed, set pressure tested, and reinstalled or replaced at least once per 18 months such that all the SRVs are tested at least once per 40 months.
This submittal proposes to revise the requirements such that only the pilot stage portions of the SRVs are required to be removed and tested at the specified frequencies, and will require the main (mechanical) stage portion of the SRVs to be set point tested at least once per 5 years.
A description of the requested amendment, supporting information and analyses for the change, and the basis for a no significant hazards consideration determination are provided in Attachment 1.
The Technical Specification pages affected by the proposed change are marked-up in Attachment 2.
Upon NRC approval of this proposed change, PSE&G requests that the amendment be made effective on the date of issuance, but implemented within sixty days to provide sufficient time for associated administrative activities.
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i MAY 211993 Document Control Desk NLR-N93054 Should you have any questions regarding this request, we will be pleased to discuss them with you.
Sincerely /
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H an ice Pr side.t'-
Nuclear Ope tions Affidavit Attachments (2)
C Mr.
T.
T. Martin, Administrator - Region I U.
S.
Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr.
S.
Dembek, Licensing Project Manager U.
S.
Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. T. P. Johnson (SO9)
USNRC Senior Resident Inspector Mr.
K. Tosch, Chief NJ Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625
REF:.NLR-N93054 LCR 92-06 STATE OF NEW JERSEY
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SS.
COUNTY OF SALEM
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l J. J. Hagan, being duly sworn according to law deposes and says:
e I am Vice Presidrat - Nuclear Operations of Public Service f
Electric and Gas Company, and as such, I find the matters set I
forth in the above referenced letter, concerning the Hope Creek Generating Station, are true to the best of my knowledge, I
information and belief.
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Subscribed and Sworn to before me this M S[
day of
/N[El
', 1993
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1.1 % ti i.J TD
< 1 MLGR No ary Public of New Jersey KIMBERLYJD Brown
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NOTARY PUBliC 0F NEWJERSEY f4 Commissim frpires April 21. legg My Commission expires on 5
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ATTACHMENT 1 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS LICENSE AMENDMENT APPLICATION SAFETY / RELIEF VALVE TESTING REQUIREMENTS FACILITY OPERATING LICENSE NPF-57 NLR-N93054 HOPE CREEK GENERATING STATION LCR 92-06 DOCKET NO. 50-354 I.
DESCRIPTION OF THE PROPOSED CHANGES As indicated on the marked-up pages in Attachment 2, PSE&G requests that:
1)
Technical Specification 4.4.2.2, Reactor Coolant System Surveillance Requirements, be revised to apply only to the pilot stage assembly portions of the Safety / Relief Valves (SRVs).
2)
A new Specification 4.4.2.3 be added to require the main (mechanical) stage portion of the SRVs to be set pressure tested at least once per 5 years.
3)
As an editorial change, Technical Specification 3.4.2.1 be revised to include a correct reference to Specification 3.4.2.2.
II.
REASON FOR THE CHANGES over pressure protection for the Hope Creek reactor coolant pressure boundary is provided by 14 Target Rock Two Stage SRVs.
The two stage design is a modification of the three stage SRV using the highly reliable first (main or mechanical) stage of the three stage SRV and replacing the less reliable second and third stages with a single, simplified pilot stage.
Technical Specifications currently require that at least half of these valves be removed, tested, and reinstalled or replaced each refueling outage.
Removal and reinstallation of these valves,.
which are located in the drywell, is a man-power intensive task.
Based upon completed work orders from the third refueling outage, each valve requires approximately 50 man-hours for removal and
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reinstallation.
This work must be performed in area radiation fields of approximately 50 mren/ hour as determined from the associated radiological work permits.
Therefore, removal and reinstallation of each SRV accounts for approximately 2.5 man-rem.
It is estimated that removal and reinstallation of only the pilot Page 1 of 5
'Attcchment 1 LCR 92-06 i
SRV Testing Requirements NLR-N93054 j
stage assemblics, in lieu of the entire SRV, would reduce the required time and dose requirements by half, thereby providing a dose savings of approximately 1.25 man-rem per valve. -This philosophy is consistent with the guidance provided in Regulatory Gdide 8.10, " Operating Philosophy For Maintaining Occupational Radiation Exposures As Low As Reasonably Achievable," which in part states:
" Modifications to operating and maintenance procedures and i
to plant equipment and facilities should be made where they will substantially reduce exposures at a reasonable cost."
In addition to the benefits obtained by reducing radiological exposure, the proposed change would also beneficially impact personnel safety.
The SRVs are situated at elevated locations within the drywell and must therefore be rigged from and to those locations during removal and reinstallation.
Each complete SRV assembly, as currently removed from the header, weighs 1100 lbs.
The pilot assemblics alone only weigh approximately 200 lbs.
Therefore, the proposed change would provide a net weight handling savings of 900 lbs per valve.
The reason for the editorial change in Technical Specification i
3.4.2.1 is to delete an incorrect reference to Specification 3.2.2, and to provide the correct reference to Specification 3.4.2.2.
III.
JUSTIFICATION FOR THE CHANGE The two stage SRV was designed to avoid spurious lifting as a result of leakage past the pilot seat, as had been experienced with the three stage SRVs.
The setpoint of the two stage SRVs is not affected by leakage in the pilot stage or in the bellows area, and thus minimizes spurious plant blowdown.
The two stage design also eliminated the problems associated with the three stage valve's pressure sensing bellows and the bellows failure sensing switch.
Design enhancements and improved valve performance notwithstanding, a generic concern of upward setpoint drift was
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experienced on some Target Rock Two Stage SRVs in the early 1980s.
A Boiling Water Reactor Owner's Group - (BWROG) sub-committee undertook an investigation to resolve this issue.
The results of the investigation identified the following two major causes that may result in upward setpoint drift (note that in both cases, the potential malfunction occurs in the pilot e
stage):
1.
Labyrinth seal induced friction due to less than the minimum required clearance between the pilot rod outside diameter and the inside diameter of the pilot liner, and t
Page 2 of 5 i
'Atttchment 1 LCR 92-06 SRV Testing Requirements NLR-N93054 2.
Corrosion induced sticking of the pilot disk to the pilot seat interface.
These results, along with recommended actions, were published by General Electric in SIL 196, Supplement 14 (April 23, 1984),
entitled " Target Rock 2 Stage SRV Set-Point Drift."
Recommendation #2 of thc SIL states:
" Refurbishment of the pilot disk and seat should be performed at least once every other outage or every three years, whichever comes first, or if as-received (*) testing indicates that a sticking pilot disk to seat condition exists....
(*) As-received tests are tests by the test laboratory prior to any valve maintenance or disassembly and following an in-service period of time."
The preceding results and recommendations were subsequently incorporated into the Hope Creek Updated Final Safety Analysis Report (UFSAR) and Technical Specifications.
Section 5.2.2.4 of the UFSAR contains a description of the equipment and components which constitute the Nuclear Pressure Relief System.
Item 5.a.3,
" Frequency of Tests and Operations" (page 5.2-24) of section 5.2.2.4.2.1.3 states:
" Operating experience at other plants has established that the HCGS can achieve optimum SRV operability by disassembly and inspection of the pilot section of a least 50 percent of the operating ve.es after each cycle.
The valves are relapped and recertified before reuse.
Every five years, elastometric seals and other environmentally sensitive materials are replaced."
Item 5.a.4,
" Service Information Letter 196 (SIL 196)," of UFSAR Section 5.2.2.4.2.1.3 (page 5.2-24) states:
" General Electric issues and maintains SIL 196 and its supplements to inform operating utilities of recommendations and product improvements that may be used to enhance valve operability, for the Target Rock two-and three-stage SRVs."
Technical Specification Surveillance Requirement 4.4.2.2 currently delineates:
"At least 1/2 of the safety relief valves shall be removed, set pressure tested and reinstalled or replaced with spares that have been previously set pressure tested and stored.in accordance with manufacturer's recommendations at least or.e.
per 18 months, and they shall be rotated such that all 14 safety relief valves are removed, set pressure tested and Page 3 of 5
' Attachment 1 LCR 93-06 SRV Testing Requirements NLR-N93054 t
I reinstalled or replaced with spares that have been i
previously set pressure tested and stored in accordance with l
manufacturer's recommendations at least once per 40 months."
J Compliance with this requirement is currently assured by removal I
of the entire SRV assemblies, including the mechanical and pilot l
valve portions.
However, the basis of the requirement, as described in the preceding discussion, applies only to the pilot valve assemblies.
Therefore, this submittal proposes to revise l
the specification to require only the pilot stage assembly i
portions of the SRVs to be removed and tested at the specified frequency.
The surveillance of the mechanical stage portions of i
the SRVs will be covered by Specification 4.4.2.3, which requires bench testing at least once every 5 years.
This surveillance i
requirement, however, is less restrictive than Technical l
Specification 4.0.5, which delineates ASME B&PV Code Section XI, IWV-3500, " Inservice Tests - Category C Valves," and requires that a predetermined percentage of the SRV population be tested at the end of each operating cycle with all SRVs be tested within 5 years (see ASME B&PV Code Section XI, Table IWV-3510-1).
An 1
exemption to this ASME B&PV Code requirement will be required, I
and until it is obtained, the surveillance frequency of the mechanical stage portion of the SRVs will be governed by i
t i
Specification 4.0.5.
PSE&G will submit a request for exemption
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l to the ASME B&PV Code surveillance requirements by June 15, 1993.
1 i
The editorial change in Technical Specification 3.4.2.1 deletes
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an incorrect reference to Specification 3.2.2, and now l
incorporates a correct reference to Specification 3.4.2.2.
1 IV.
Sionificant Hazards Consideration Evaluation a
PSE&G has, pursuant to 10 CFR 50.92, reviewed the-proposed amendment to determine whether our request involves a significant hazards consideration.
We have determined that operation of the i
Hope Creek Generating Station in accordance with the proposed changes:
1 1.
Will not involve a significant increase in the probability or consequences of an accident previously evaluated.
Technical Specifications (TSs) currently require that one-half of the safety / relief valves (SRVs) be tested at least once per 18 months, and that they be rotated such that all 14 SRVs are tested at least once per 40 months.
This requirement was incorporated into the Hope Creek TSs based upon the recommendations of General Electric Service Information Letter (SIL) 196, which was issued to address l
j concerns relative to upward setpoint drift resulting from potential malfunctions in the pilot stage of the SRVs.
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PSE&G believes that the mechanical stage of the SRVs has J
Page 4 of 5 A
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Attechnent 1 LCR 92-06 SRV Testing Requirements NLR-N93054 proven to be highly reliable and need not be subject to these requirements.
We are therefore proposing that the mechanical stage portion of the SRVs be tested at least once every 5 years.
The pilot stage of the SRVs will continue to be tested in accordance with the recommendations of SIL 196.
PSE&G believes that these proposed testing requirements will not significantly affect the reliability of the SRVs and will continue to ensure adequate capability of the SRVs to perform their intended safety functions.
We therefore i
believe that the proposed changes will not significantly increase the probability or consequences of a previously analyzed accident.
2.
Will not create the possibility of a new or different kind I
of accident from any accident previously evaluated.
This proposal does not involve any hardware or logic j
changes, nor alters the way in which any plant system is operated; therefore, there are no new possibilities or types of accidents introduced.
3.
Will not involve a significant reduction in a margin of safety.
As discussed in Criterion 1 above, the proposed testing I
frequency and applicability will provide a comparable degree of assurance that the SRVs will be capable of performing their intended function.
In addition, the implementation of the proposed amendment will result in a reduction of radiological exposure of plant personnel and provide an enhancement to personnel safety.
We therefore believe that the proposed change will not significantly reduce a margin of safety.
i V.
Conclusion i
Based on the preceding discussion, PSE&G has concluded that the proposed change to the Technical Specifications does not involve a significant hasards consideration insofar as the change:
(i) does not involve a significant increase in the probability or consequences of an accident previously evaluated, (ii) does not create the possibility of a new or different kind of accident from any accident previously evaluated, and (iii) does not involve a significant reduction in the margin of safety.
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