ML20073L761
| ML20073L761 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 09/28/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20073L758 | List: |
| References | |
| NUDOCS 9410130281 | |
| Download: ML20073L761 (9) | |
Text
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UNITED STATES f
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20566-0001
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4 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 217 TO FACILITY OPERATING LICENSE NO. DPR-59 POWER AUTHORITY OF THE STATE OF NEW YORK i
JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 4
1.0 INTRODUCTION
By letter dated December 20, 1989, as supplemented by letters dated January 16, 1990, January 3, 1992, January 30, 1992, May 5, 1993, and May 26, J
1993, and superseded March 2,1994, the Power Authority of the State of New York (the licensee) submitted a request for changes to the James A.
FitzPatrick Nuclear Power Plant Technical Specifications-(TSs). The requested changes would modify the Safety / Relief Valve (SRV) performance limits. Specifically, the requested changes would:
(1) modify TS 2.2.1.B.
and ~its associated Bases, to establish a single nominal SRV setpoint of 1110 psig; (2) modify TS 4.6.E, and its associated Bases, to. increase the SRV setpoint tolerance to 3%; and (3) codify TSs 3.5.0, 3.6.E, and 4.5.D, and their associated Bases, to allow for two SRVs (or Automatic Depressurization System (ADS) valves) to be inoperable during continuous power operation.
In addition, miscellaneous changes were requested to several TSs to clarify terminology, correct typographical errors, remove a surveillance requirement which should have been deleted as part of Amendment No. 130, and delete a duplicate TS. Specifically, the requested changes would:
(1) modify TS 1.2.1, and the Bases sections for TSs 3.6.E and 4.6.E, to clarify terminology; (2) modify TS 3.5.D.2, and the Bases sections for TSs 1.2 and 2.2, to correct typographical errors; (3) modify TS 4.2.B, Table 4.2-2, to correct an error made in Amendment No. 130 that failed to dalete the requirement to perform logic functional testing on the ADS bellows pressure switch; (4) modify TS 4.5.D.1.b, to move the TS to 4.6.E.4, a new sectian, and clarify the requirements associated with SRV manual actuation testing; (5) modify Bases sections for TSs 3.6.E and 4.6.E, to move the bases for the SRV manual actuation testing to the applicable sections; and (6) modify TS 4.6.E.4, to delete a duplicate TS pertaining to the annual report of SRV failures. The miscellaneous proposed changes are administrative in nature.
9410130281 940928 PDR ADOCK 05000333 P
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2.0 DISCUSSION The James A. FitzPatrick Nuclear Power Plant is designed with eleven SRVs, seven of which are also ADS valves.
Technical Specification 2.2.1.B currently requires the SRV nominal settings to be 1090 psig for two SRVs, 1105 psig for two SRVs and 1140 psig for the i
remaining seven SRVs.
TS 4.6.E currently allows a setpoint error of +/- 1%
l for each SRV. The TSs assure that the structural acceptance criteria set forth in the Mark I containment Short Term Program are satisfied.
TS 3.5.0 currently requires all seven ADS valves to be operable during plant operation, with a contingency that plant operation can continue for 30 days with one ADS valve inoperable. The redundancy requirement assures adequate core spray and low pressure coolant injection flow in the event of a small pipe break coincident with a failure of the high pressure coolant injection system.
To reduce the number of forced outages and decrease maintenance and surveillance testing costs, the licensee has proposed:
(1) to modify l
TS 2.2.1.8 to establish a single nominal SRV setpoint of 1110 psig; (2) to modify TS 4.6.E, to increase the SRV setpoint tolerance to 3%; and (3) to modify TS 3.5.D to permit two ADS valves or SRVs to be inoperable whenever the reactor pressure is greater than 100 psig and irrauiated fuel is in the reactor vessel, and prior to reactor startup from a cold shutdown.
In addition, modifications to TS 3.6.E and to Bases Sections 2.2.1.B, 3.6.E, and 4.6.E, were proposed to support the charges proposed to TSs 2.2.1.8, 4.6.E, and 3.5.D.
The licensee has also proposed modifications, all administrative in nature, to several TSs associated with the SRVs. The following is a list of all the proposed changes:
1.
TS 1.2.1: change the phrase " reactor coolant system pressure" to
" reactor vessel dome pressure."
2.
TS 2.2.1.B:
change
" Reactor coolant system safety / relief valve nominal settings shall be as follows:
Safety / Relief Valves 2 valves at 1090 psig 2 valves at 1105 psig 7 valves at 1140 psig l
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The allowable setpoint error for each safety / relief valve shall be +/- 1 percent."
to reed:
"At least 9 of the 11 reactor coolant system safety / relief valves shall have a nominal setting of 1110 psig with an allowable setpoint error of i 3 percent."
3.
TS 3.5.D.1:
replace with the following:
"The ADS shall be operable with at least 5 of 7 ADS valves operable:
a.
wheaever the reactor pressure is greater than 100 psig and irradiated fuel is in the reactor vessel, and b.
prior to reactor startup from a cold condition."
4.
TS 3.5.D.2: delete the "," after "100 psig."
l 5.
TS 3.5.D.3: delete the cross-reference to action statements 3.5.D.1.a j
and 3.5.D.1.b and add " ADS." The revised TS reads:
" Low power physics testing ana reactor operator training shall be permitted with inoperable ADS components, provided that reactor coolant temperature is 5; 212 *F and the reactor vessel is vented or reactor vessel head is removed."
l 6.
TS 3.6.E.1: delete the words " Safety and" from the title, change the word "all" to "at least 9 of 11," and delete the phrase "except as specified by Specification 3.6.E.2."
The revised TS shall read as follows:
"During reactor power operation conditions and prior to startup from a cold condition, or whenever reactor coolant pressure is greater than atmospheric and temperature greater than 212 'F, the safety mode of at least 9 of 11 safety / relief valves shall be operable. The Automatic l
Depressurization System valves shall be operable as required by l
specification 3.5.D."
7.
TS 3.6.E.2: delete specification.
8.
TS 3.6.E.3:
delete the cross-reference to TS 3.6.E.2 and renumber the TS to be 3.6.E.2.
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TS 3.6.E.4:
change the cross-reference from " Item B.2" to " Technical Specification 3.6.E.1" and renumber this specification to be 3.6.E.3.
- 10. TS 3.6.E.4 (New):
add "The provisions of Technical Specification 3.0.0 l
are not applicable."
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Switch."
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- 12. TS 4.5.D.1.b:
move this specification to new Section 4.6.E.4 and add "This test shall be performed at least once each operating cycle within the first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of continuous operation at a reactor steam dome j
pressure of ;t 940 psig."
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- 13. TS 4.5.D.2-revise to read as follows:
"A logic system functional test.
a.
When it is determined that two valves of the ADS are inoperable, the ADS subsystem actuation logic for the operable ADS valves and the HPCI subsystem shall be verified to be operable immediately and at least weekly thereafter.
b.
When it is determined that more than two relief / safety valves of the ADS are inoperable, the HPCI System shall be verified to be operable immediately."
- 14. TS 4.6.E.1: delete the words " Safety and" from the title, change "one half of all" to "5 of the 11," delete cross-reference to TS 2.2.B, and add the revised valve actuation setpoints. The revised TS shall read as follows:
"At least 5 of 11 safety / relief valves shall be bench checked or replaced with bench checked valves once each operating cycle. All valves shall be tested every two operating cycles. The testing shall demonstrate that the 11 safety / relief valves actuate at 1110 psig.+/- 3%."
- 15. TS 4.6.E.4: delete the following:
"An annual report of safety / relief valve failures and challenges will be sent to the NRC in accordance with Section 6.9.A.2.b."
In addition to the proposed modifications to the TS associated with the SRVs, the licensee has proposed changes to the associated the Bases sections.
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. j 3.0 EVALUATION The licensee has proposed several changes, both technical and administrative in nature, to the TSs associated with the SRVs.
First, the licensee proposed technical changes to establish:
(1) a single i
nominal SRV setpoint of 1110 psig; (2) a SRV setpoint tolerance of 3%; and (3) an allowance for two SRVs to be inoperable during continuous power operation.
To justify the change, the licensee submitted two reports prepared by General Electric (GE): NEDC-31967P-Revision 1 and OPE 10-379. The reports l
present the results of licensing bases calculations showing that with the
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proposed modifications, vessel overpressurization limits and loss-of-coolant accident / emergency core cooling system (LOCA/ECCS) performance requirements are satisfied. The reports also show that the proposed changes do not have a significant impact on plant piping and containment structure performance. At the NRC staff's request, additional information on plant response to Anticipated Transient Without Scram (ATWS) events (NEDE-24222, " Assessment of BWR Hitigation of ATWS", Volume II), was provided by the licensee. The report shows that in the event of the most limiting ATWS event, the peak reactor vessel pressure will be within acceptable limits.
The following provides the details of the staff's review of the proposed changes:
1 Effects Of Proposed Chances On Reactor Vessel And ECCS
- The licensee's proposed SRV nominal setpoint (1110 psig) is set below the reactor vessel design pressure of 1250 psig, satisfying the requirements of Article 9 of the American Society of Mechanical Engineers Boiler and.
Pressure Vessel Code-Section III, Nuclear Vessels. The setpoint is low enough to ensure high-pressure coolant injection and reactor core isolation coolant rated flow is still achievable and turbine overspeed margins are maintained. Also, since the proposed setpoint is within the range of the i
current staggered setpoints, the overall likelihood of inadvertent valve opening (from either downward setpoint drift or simmer problems) is not
.'xpected to change significantly.
The staff finds the proposed SRV nominal setpoint and tolerance change has no significant impact on the reactor vessel and ECCS operation.
- NEDE-24222 concluded that for the most limiting ATWS event (main steam isolation valve closure, with all SRVs operable and Alternate Rod Insertion system failure) a peak reactor vessel pressure of 1296 psig will be reached. With two SRVs out-of-service, the nominal relief valve capacity will be reduced by approximately 18 percent. As a result, peak vessel pressure will increase by 137 psi to approximately 1433 psig.
Since the peak vessel pressure is less than the peak pressure expected to occur under 4
the most limiting ATWS event (1500 psig), the staff finds operation of the plant with two SRVs inoperable to be acceptable.
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I Effects Of Proposed Chances On Plant Pipino And Containment Structures
- The staff reviewed the licensee's supporting technical evaluations which considered the effects of increased loads that will result from the i
simultaneous actuation of all SRVs. These evaluations analyzed the following loads:
(a)
SRV thrust loads on the main steam piping.
(b)
Loads on SRV discharge piping due to increased motion of main steam lines.
(c)
Water jet loads on submerged structures, l
(d)
Air bubble drag loads.
(e)
Torus shell loads.
(f)
Torus support loads.
(g)
Torus attached piping loads.
The plant main steam lines were analyzed in OPE 10-379 for the simultaneous actuation of all SRVs at 1140 psig. The analysis considered the increased I
motion of these lines and th?. attached SRV discharge piping by analyzing the steam line which the licensee determined is the most highly stressed.
1 The licensee stated that the FitzPatrick Mark I Plant Unique Analysis l
Report (PVAR) calculated the effects of the simultaneous actuation of all l
SRVs at 1140 psig on the submerged structures, torus shell, torus supports and attached piping. This analysis demonstrates that the resulting loads will not cause the stresses in these components to exceed the allowable values. These stresses were determined by combining the SRV discharge loads with other appropriate loads including the safe shutdown earthquake loads.
Thrust loads on SRV piping and T-quenchers were determined using the relief valve forced outage rate (RVFOR) computer program.
After the analysis was performed, the licensee discovered an error in an equation which determines the water clearing thrust loads and associated stresses in the submerged SRV discharge piping. The effects of this error have been studied by the licensee, and the results of the tests at the Monticello plant were compared with load; predicted by a version of RVFOR with the error and a corrected version of RVFOR. These comparisons show that Monticello's test results are consistent with the loads predicted by the corrected version of RVFOR and that the version with the error overestimates these loads by 40%
to 50%. A letter from GE to the licensee, dated May 25, 1984, indicates that the reduction of these loads is not constant and should be quantified by plant unique calculations if credit for their reduction is to be taken.
Tests were also conducted at FitzPatrick and the resulting loads were compared to the RVFOR model results.
As a result, it was determined that the loads predicted by RVFOR with the error are conservative for i
er
FitzPatrick, and there is sufficient margin in these components to withstand the water clearing thrust loads at FitzPatrick as analyzed in the PUAR.
Based on the analysis performed for the plant piping and containment structures for the simultaneous discharge of all 11 SRVs at 1110 psig, the licensee has determined that the allowable stresses in these plant components will not be exceeded for the limiting combination of loads.
Based on the evaluation, the staff agrees that the analysis which the licensee has provided demonstrates the adequacy of the plant piping and containment structures for the proposed SRV setpoint and tolerance change.
The licensee has shown that for the simultaneous discharge of all 11 SRVs 4
at a single nominal setpoint of 1110 psig, the allowable stresses in these plant components will not be exceeded. The staff finds the proposed TS SRV setpoint change has no significant safety impact on plant piping and j
containment structures.
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- A review of the original analysis on the impact of the increased SRV pressure setpoint on containment response shows that the setpoint used in the analysis was slightly higher than the proposed value.
In addition, the most limiting drywell pressure transient was found to be the design basis LOCA rather than a SRV event, and the most limiting drywell temperature transient was found to be the main steam line break event. The licensee has determined that neither of these transients is affected by the increased SRV setpoint.
The licensee's containment response analysis shows that the increased SRV setpoint is bounded by the original analyses. However, the proposed increase in setpoint tolerance will cause the maximum possible relief pressure to exceed the analyzed worst case by approximately 3 psi.
Based on the assumptions used in determining the blowdown mass and energy release rate, the slight overshoot is considered to be insignificant relative to the containment temperature and pressure response. The staff concludes that the licensee's analysis of containment temperature and pressure response remains as the bounding analyses and is acceptable given the extremely small change in blowdown mass and energy input into the drywell via the SRVs.
Therefore, the staff finds that the proposed TS SRV setpoint change has no significant impact on the containment.
- The licensee evaluated the effects of the increased SRV setpoint and tolerance on the containment hydrodynamic loads.
The loads, documented in the PVAR, were applied to the containment and were found to Le most affected by the fluid mass rate from the SRV.
The PUAR assumed a mass flow rate of 291 lb/sec for each SRV at a setpoint of 1140 psig. The PUAR assumed that SRV load case A 1.2 and C 3.2 were the bounding cases with steam flow rates of 291 lbs/sec.
The resultant hydrodynamic loads, which 4
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occur from an SRV discharge, are essentially proportional to the mass flow rate from the SRV.
The licensee has addressed the effect of increased SRV setpoints and the tolerance change upon the analyzed SRV flows and determined that the maximum flow rate is bounded by the assumed value of 291 lbs/sec. The staff finds the licensee's analysis acceptable based on the determination that the bounding case steam flow rates used in the PUAR are not exceeded by the proposed SRV setpoint change.
In addition to the proposed technical changes, the licensee proposed several administrative changes to the TSs, unrelated to SRV performance, to correct typographical errors, to remove a surveillance requirement which should have been deleted as part of Amendment No.130, to clarify when SRV manual actuation is performed, and to delete a duplicate TS. The staff finds the changes:
(1) do not involve a plant modification, (2) do not impact any procedural or administrative controls, and (3) do not involve a reduction in safety margin.
Therefore, the staff finds the changes to be acceptable.
Based on the above evaluation, the staff concludes that there is no significant safety impact on vessel overpressure margin,.ECCS/LOCA performance, plant piping, or containment structures due to operation with:
(1) two SRVs out of service, (2) all 11 SRV setpoints set at a single nominal setpoint of 1110 psig, and (3) a setpoint tolerance of +/- 3%. Therefore, the staff finds these proposed changes, in addition to the miscellaneous changes which are administrative in nature, to be acceptable. The staff has no objections to the proposed changes to the TS Bases.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32 and 51.35, an environmental assessment and finding of no significant impact was published in the Federal Reaister on July 14, 1993 (58 FR 37972). By letter dated March 2, 1994, the licensee subsequently superseded the original application for amendment as supplemented. The application dated March 2,1994, differed from the original application in that it eliminated a provision in the original application that would have allowed the use of an SRV setpoint upper limit as an acceptable means to reduce the number of License Event Reports.
The environmental assessment and finding of no significant impact published on July 14, 1993, therefore, addressed all of the provisions of the application dated March 2, 1994.
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j Accordingly, based upon the environmental as'sessment, the Comission has determined that issuance of this amendment will not have a significant effect 4
on the quality of the human environment.
6.0 CONCLUSION
The Comission has concluded, based on the considerations discussed above, i
that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such j
activities will be conducted in compliance with the Comission's regulations, and (3) the issuance of the amendment will not be inimical to the comon defense and security or to the health and safety of the public.
Principal Contributors:
C. Hamer M. Razzaque i
A. D'Angelo Date: September 28, 1994
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