ML20129A708
| ML20129A708 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 09/11/1996 |
| From: | VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | |
| Shared Package | |
| ML20129A693 | List: |
| References | |
| NUDOCS 9609190030 | |
| Download: ML20129A708 (34) | |
Text
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k ATTACHMENT A
, Proposed Change #185 to Technical Specifications and Bases
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} Affected Pages
- Technical Specification page 18 ,
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- Technical Specification page 120 i
- Bases page 142 J
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9609190030 960911 PDR ADOCK 05000271 P PDR
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l VYNPS 1.2 SAFE"'Y LIMIT 2.2 LIMITING SAFE?t SYSTEM SE7:':NC
! l 1.2 REAC'"0R COOLAIC SYSTm 2.2 REACMR COCLAfC SYSTO Acellesbilitv: Aceliesbiliev: ,
Applies to limits on reactor Applies to trip settings for coolant system pressure. controlling react:r system pressure.
Ob4ectives Obiectives Tc establish a limit below which To provide for pr:tectiva metien
. the integrity of the reactor in the event that the 6 r in c ip @
coolant system is not threatened process variable approacnes a due to an overpressure safety limit.
j condition.
Someifieseion: Seecifiestten: p et a Cif M The reactor coolant system pressure shall not exceed 1335 psig at any time when
, irradiated fuel is present in the reactor vessel.
A. Reactor coolant high pressure scram shall be less '
- than or equal to 1055 psig.
1 j B. Primary systa:n relief and safety valve settings shall f speedied & TGA 2.21 bea=gp'*=>
va' e at <* 80 p ig
. v lv s a <1090 psi 1 <a'<e a 5.110 psi v lves at <l's 0 p ig ety 7 .ves
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TABLE 2.2.1 m/
poIVARv svsm RELITF AND 9 aft"'"! VAiW SENINO2, Number and Type of Valve (s) Lift Setting W 1 safety relief valve 1080 psig .
2 safety relief valves 1090 psig i safety relief valve 1100 psig 2 safety valves 1:40 psig
?tble 2.1.1 Nete t
1, As-left setpoint tolerance i1 %
As-found setpoint tolerance 3% .
Amendment No. W 18
VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS CPEPATION l Safety and Relief valves D. Safety and Relief Valves D.
- 1. During reactor power 1. Operability testing of ,
operating conditions and Safety and Relief Valves whenever the reactor shall be in accordance coolant pressure is with Specification greater than 120 psig 4.6.E. The lift point and temperature greater of the safety and relief than 3500F, both safety valves shall be set as valvesjsna i ou specified in o racle. The elief Specification 2.2.B.
Ives all opera e, ex ept t e if one lief alve in erabl , reac or gnd gh \ eag3}. 3 of 4he Q p er sh 1 be
.mmedi ely r uced o (b h e. ( V a.IV e $. Sh m.ll lo e.
and intai d at o bel 95% rate O )D*C" b 8
- o er.
- 2. If Specification 3.6.D.1 l
is not met, initiate an I orderly shutdown and the reactor coolant pressure shall be below 120 psig and 3500F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
E. Stractural Intecrity and E. Structural Intecrity and l Ooerability Testino operability Testino The structural integrity and 1. Inservice inspection of the operability of the safety-related safety-related systems and components shall be components shall be performed in accordance maintained at the level with Section XI of the required by the original ASME Boiler and Pressure acceptance standards Vessel Code and throughout the life of the applicable Addenda as plant. required by 10 CFR 50, Section 50.55a(g),
except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g) (6) (i) .
Inservice inspection of piping, identified in NRC Generic
- Letter 88-01, shall be performed in accordance with the staff positions l on schedule, methods, j and personnel and sample l expansion included in j the Generic Letter. i 1
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Amendment No. 44, te, 46, 99, 4G4, 499- 120
VYNPS j BASES: 3.6 and 4.6 (Cont'd) impurities will also be within their normal ranges. The reactor cooling samples will also be used to determine the chlorides.
Therefore, the sampling frequency is considered ' adequate to detect j leng-term changes in the chloride ion content. Isot:pic analyses i required by Specification 4.6.B.2 may be performed by a gam:na scan i and gross beta and alpha determination. i l
The conductivity of the feedwater is continuously menitored and alarm i set points consistent with Regulatory requirements given in Regulatory cuide 1.56,
- Maintenance of Water Purity in Boiling Water Reactors,* have been determined. The results from the conductivity monitors on the feedwater can be correlated with the results from the conductivity monitors on the reactor coolant water to indicate dominerali:er breakthrough and subsequent conductivity levels in the reactor vessel water. l C. Coolant Leskace The 5 gpm limit for unidentified leaks was established assuming such l 1eakage was coming from the reactor coolant system. Tests have been )
conducted which demonstrate that a relationship exists between the j size of a crack and the probability that the crack will propagate. l These tests suggest that for leakage somewhat greater than the limit specified for unidentified leakage; the probability is small that imperf ections or cracks associated with such leakage would grow rapidly. Leakage less than the limit specified can be detected within a few hours utilizing the available leakage detection systems.
If the limit is exceeded and the origin cannot be determined in a reasonably short time the plant should be shutdown to allow further investigation and corrective action.
The 2 gym increase limit in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period for unidentified leaks was established as an additional requirement to the 5 gpm limit by Generic Letter 88-01, 'NRC Position on Intergranular Stress Corrosion Cracking (ICSCC) in BWR Austenitic Stainless Steel Piping.
- j The removal capacity from the drywell floor drain su ; and the equivalent drain sump is 50 gpm each. Removal of 50 gpm f rom either j of these sumps can be accomplished with considerable margin. J D. Safetv and Relief valves Paramet ic evaluat'ons have she that only hree of the f r relie l valve are requi- d to provide a pressure rgin greater an the reco ended 25 i below the afety valve etuation sett gs as well) as MCPR > 1. 6 for the li ting overpr sure transien below 981 )
pc er. Cons ently, 95% wer has bee selected as a imiting pc er }
vel for t .ee valve ope ation. For . e purp .;es of this limit g ondition relief valv that is unab e to actaate . thin toler ce of its a press tre is onsidered to be as inopera e as a mechani lly malt act ning valve.
Exper ence in safet valve opera on shows that a testing o 50% of the afety valves er refueling utage is ade ate to dete : failures or eterioration. The toleran e value is spe ified in Se tion III of tr ASME Boiler nd Pressure essel Code as it of desig pressure.
analysis ha been perform d which shows at with al safety alves set it igher the re ctor coolant p essure safe limit of 1375 psig is not exceedeg ,
.See L sert Ch....s 10nt uh 20, 1974, H, M, We, 1:0 142
Insert for BASES 3.6 and 4.6.D Safety and Relief Valves Safety analyses have shown that only three of the four relief valves are required to provide the recommended pressure margin of 25 psi below the safety valve actuation settings as well as compliance with the MCPR safety limit for the limiting anticipated overpressure transient. For the purposes of this limiting condition, a relief valve that is unable to actuate within tolerance of its set pressure is considered to be as inoperable as a mechanically malfunctioning valve.
The setpoint tolerance value for as-left or refurbished valves is specified in Section III ,
of the ASME Boiler and Pressure Vessel Code as il% of set pmssure. However, the I code allows a larger tolerance value for the as-found condition if the supporting design analyses demonstrate that the applicable acceptance criteria are met. Safety analysis has been performed which shows that with safety and safety relief valves within 3% of the specified set pressures in Table 2.2.1 and with one inoperable safety relief valve, the reactor coolant pressure safety limit of 1375 psig and the MCPR safety limit are not !
exceeded during the limiting overpressure transient.
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- 0 ATTACHMENT B Proposed Change #185 to Technical Specifications and Bases New Pages
- Technical Specification page 18
- Technical Specification page 120
- Bases page 142 P
VYNPS 1.2 SAFETY LIMIT 2.2 LIMITING SAFETY SYSTEM SETTING 1.2 REACTOR COOLANT SYSTEM 2.2 REACTOR COOLANT SYSTEM l Apolicability: Aoplicability:
I I Applies to limits on reactor Applies to trip settings for l coolant system pressure, controlling reactor system
- pressure.
Obiective: Obiective:
To establish a limit below which To provide for protective action the integrity of the reactor in the event that the principal l coolant system is not threatened process variable approaches a i due to an overpressure safety limit.
I condition.
Specification: Specification:
f- The reactor coolant system pressure shall not exceed 1335 psig at any time when irradiated fuel is present in the reactor vessel.
A. Reactor coolant high pressure scram shall be less than or equal to 1055 psig.
B. Primary system relief and safety valve settings shall be as specified in Table 2.2.1.
TABLE 2.2.1 Primary Svstem Relief and Safetv Valve Settinos Number and Type k.if t ,
of valve (s) settingm -l 1 safety relief valve 1080 psig 2 safety relief valves 1090 psig 1 safety relief valve 1100 psia 2 safety valves 1240 psig Notes (1) As-left setpoint tolerance 21%.
As-found setpoint tolerance 23%.
Amendment No. H, 18
VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION D. Safety and Relief Valves D. Safety and Relief Valves
- 1. During reactor power 1. Operability testing of operating conditions and Safety and Relief Valves whenever the reactor shall be in accordance coolant pressure is with Specification greater than 120 psig 4.6.E. The lift point and temperature greater of the safety and relief than 350oF, both safety valves shall be set as valves and at least specified in three of the four relief Specification 2.2.B.
valves shall be operable.
- 2. If Specification 3.6.D.1 is not met, initiate an orderly shutdown and the reactor coolant pressure shall be below 120 psig and 350oF within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
E. Structural Integrity and E. Structural Integrity and Operability Testing Operability Testing The structural integrity and 1. Inservice inspection of the operability of the safety-related safety-related systems and components shall be components shall be performed in accordance ;
maintained at the level with Section XI of the l required by the original ASME Boiler and Pressure l acceptance standards Vessel Code and i throughout the life of the applicable Aldenda as plant, required by 10 CFR 50, Section 50.55a(g), l except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
Inservice inspection of piping, identified in NRC Generic Letter 88-01, shall be performed in accordance with the staff positions on schedule, methods, and personnel and sample expansion included in the Generic Letter.
Amendment No. 13, 44, 46, 99, +G4, 439, 120
VYNPS BASES: 3.6 and 4.6 (Cont'd) I impurities will also be within their normal ranges. The reactor cooling samples will also be used to determine the chloridos.
Therefore, the sampling frequency is considered adequate to detect long-term changes in the chloride ion content. Isotopic analyses required by Specification 4.6.B.2 may be performed by a gamma scan and gross beta and alpha determination. l The conductivity of the feedwater is continuously monitored and alarm set points consistent with Regulatory requirements given in j Regulatory Guide 1.56, "Mairtenance of Water Purity in Boiling Water -
Reactors," have been determined. The results from the conductivity monitors on the feedwater can be correlated with the results from the conductivity monitors on the reactor coolant water to indicate demineralizer breakthrough and subsequent conductivity levels in the )
reactor vessel water.
C. Coolant Leakage l The 5 gpm limit for unidentified leaks was established assuming such leakage was coming from the reactor coolant system. Tests have been conducted which demonstrate that a relationship exists between the size of a crack and the probability that the crack will propagate. i These tests suggest that for leakage somewhat greater than the limit l specified for unidentified leakage; the probability is small that j imperfections or cracks associated with such leakage would grow l rapidly. Leakage less than the limit specified can be detected '
within a few hours utilizing the available leakage detection systems.
If the limit is exceeded and the origin cannot be determined in a reasonably short time the plant should be shutdown to allow further l investigation and corrective action.
The 2 gpm increase limit in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period for unidentified leaks was established as an additional requirement to the 5 gpm limit by Generic Letter 88-01, "NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping."
The removal capacity from the drywell floor drain sump and the equivalent drain sump is 50 gpm each. Removal of 50 gpm frem either of these sumps can be accomplished with considerable margin.
D. Safety and Relief Valves Safety analyses have shown that only three of the four reljef valves are required to provide the recommended pressure margin of 25 psi below the safety valve actuation settings as well as compliance with the MCPR safety limit for the limiting anticipated overpressure transient. For the purposes of this limiting condition, a relief valve that is unable to actuate within tolerance of its set pressure is considered to be as inoperable as a mechanically malfunctioning valve.
The setpoint tolerance value for as-left or refurbished valves is specified in Section III of the ASME Boiler and Pressure Vessel Code as 11% of set pressure. However, the code allows a larger tolerance value for the as-found condition if the supporting design analyses demonstrate that the applicable acceptance criteria are met. Safety analysis has been performed which shows that with all safety and safety relief valves within 13% of the specified set pressures in Table 2.2.1 and with one inoperable safety relief valve, the reactor coolant pressure safety limit of 1375 psig and the MCPR safety limit are not exceeded during the limiting overpressure transient.
Changc 15/ March 20, 1970, 14, 14, 196, +}9, 142
ATTACHMENT C Proposed Change #185 Justification For Operation With A Relaxed Safety Valve Setpoint Tolerance and 100% of Rated Power Operation With An Inoperable SRV For The Vennont Yarikee Nuclear Power Station 4
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1.0 EXECtfTIVE StiMM A RY I
1his repoit (kouments safety analyses perfonned in support of the Vermont Yankee Nuclear Power Station. The analyses were performed to support implementation of the following two changes:
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- 1) The setpoint tolerance of safety relief valves and safety valves may be increased from *1% to *3%. As.round valve setpoints within i3% of Technical Specification setpoints are acceptable.
- 2) Technical Specifications cunently require that reactor power be maintained s 95% of rated thennal power when one of the four safety relief valves is inoperable. The supporting safety analyses demonstrate that this restnetion ;
is not necessary to meet the applicable acceptance criteria.
Justification for these changes is provided by presenting the results of safety analysis for those postulated events which may challenge the safety reliefvalves or safety valves. Safety analysis results demonstrate that applicable acceptance criteria are met for the above changes.
Section 2 of this report describes the scope of the safety analyses performed, the related acceptance criteria, and the computer codes used. The applicable acceptance criteria for the changes include:
e transient reactor pressure vessel pressure s ASME Boiler Vessel Code limit (110% of design) e no significant increase in safety relief valve or safety valve challenges e fuel integrity assured by MCPR 2 the safety limit value during overpressure transients
- LOCA design basis criteria for emergency core cooling systems, contaimnent design basis, containment heat removal, radiological releases, and LOCA induced stnictural loads e associated stnictures, systems, and components remain intact during safety relief valve discharge Section 3 describes the analyses of postulated overpressure events and the results. The discussion includes a description of the SRV and SV modeling changes required to simulate the effects of a increased setpoint tolerance and an inoperable SRV. The analyzed cases are organized by the acceptance criteria which they support.
Section 4 describes the Loss of Coolant Accident (LOCA) analysis and the results.
Section 5 provides a mechanicalloads analysis performed to verify the continued integrity of the SRV discharge piping and Tonis during operation with a increased setpoint tolerance for all SRVs.
Section 6 provides the conclusions derived from the results of analyses. The above two changes may be implemented separately or in combination. Safety analysis results demonstrate that all the applicable acceptance criteria are met.
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2.0 INTRODITCTION l
2.1 Scone of Analnes The Vermont Yankee reactor vessel safety / relief conhguration includes 4 Safety Relief Valves (SRVs) which discharge to the Torus and 2 Safety Valves (SVs) w hich discharge to the drywell. The current Technical Specification setpoints are 1 SRV @ 1080 psig 2 SRVs @ 1090 psig,1 SRV @ l 100 psig, and 2 SVs @ 1240 psig. The current setpoint tolerance on all these valves is *1% as renuired by the ASME Boiler and Pressure Vessel Code. These setpoints and their relationsidp to other relevant plant pressure parameters are shown in Figure 2-1. The proposed changes to the configuration include:
- 1) A change to alkyw as-found setpoints for the SRVs and SVs to be *3% of the Technical Specification values. The current practice of meeting *1% tolerance for the as-left condition as required by the ASME Boiler and Pressure Vessel C(xle would continue. The range of valve setpoints with *3% tolerance is also shown in Figure 2-1.
- 2) A change to the Technical Specifications to remove the current requirement to operate at c 95% rated power with an inoperable SRV. Operation at full power with an inoperable SRV would be permitted.
1 Safety analyses are performed to demonstrate that operation with these changes will not result in violation of applicable )
acceptance criteria. The safety analyses include- ,
a) overpressure transient analysis to demonstrate that the ASME overpressure limit (110% of design) continues to be met. This analysis is discussed in Section 3.2, ,
1 b) overpressure transient analysis to demonstrate no significant increase in challenges to SRVs and SVs. This analysis is discussed in Section 3.4; !
c) a hot channel analysis to detennine the impact of the changes on the operating Minimum Critical Power Ratio (MCPR) limits specified in the Core Operating Limits Repod (COLR). This analysis is discussed in Section 3.3; d) less of Coolant Accident (LOCA) analyses to demonstrate continuing compliance to LOCA design basis criteria for emergency core cooling systems, containment design basis, containment heat removal, radiological releases, and LOCA induced structural loads. This analysis is discussed in Section 4; and c) mechanical loads analysis of the SRV discharge piping and Torus to demonstrate they remain intact during SRV discharge. This analysis is discussed in Section 5.
2.2 Acceptance Criteria and Their Annlicability 1he following set of acceptance criteria are applied in the safety analysis of the increased setpoint tolerances and full power operation with an inoperable SRV.
2.2.1 AS. ME Ovemressure Limit l l
The ASME Boiler and Pressure Vessel Code Section III-A permits pressure transients up to 10% over design pressure (110% x 1250 = 1375 nsig). The limiting overpressure Abnormal Operational Transient (AOT) analyzed is a Main Steam Isolation Vahe (MSIV) ckwure at End of Full Power Life (EOFPL) without credit for reactor trip on MSIV position sensing.
The next most limiting overpressure AOT is the Generator Load Rejection Without Bypass (GLRWOBP) at EOFPL. These two transients are re-analyzed in this repon to demonstrate continued compliance with the ASME overpressure limit (110%
of design). Specifically, the maximum transient pressure in the bottom of the reactor vessel must be s 1375 psig.
2.2.2 Operating MCPR Limits Operational restraints on the minimum critical power ratio (MCPR) are placed in the Core Operating Limits Report (COLR)
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1 to assure that safety limits on CPR are not violated during AOTs. The impact of the changes described in Section 2.1 on MCPR limits is detennined by performing a hot channel analysis for the overpressure transient which yields the largest transient drop in CPR (ACPR). This transient is typically the GLRWOBP. The analysis is discussed in Section 3.3.
2.2.3 SRV and SV Challenges in addition to the acceptance criteria on maximum overpressure (see Section 2.2.1) and MCPR (see Section 2.2.2), other operational restramts are imposed to provide assurance that SRV and SV challenges are minimized.
From Tecimical Specification 3.6 D.lm, reactor power cunently must be maintained at or below 95% rated power when one of the four SRVs is inoperable. This requirement was established to assure that SVs and fuel integrity are not challenged during AOTs while operating with an inoperable SRV. Since operating MCPR limits are established to assure fuel integrity during AOTs, fuel integrity is already assured for all AOTs. Since SVs discharge to the drywell and the discharge may contain low levels of radioactivity , it is pmdent to minimize SV challenges. In Section 3.4, overpressure transient analysis is performed with the changes discussed in Section 2.1 to demonstrate challenges to SVs are unlikey during AOTs while operating at full power with an inoperable SRV. For this purpose, a GLRWOBP is the limiting AOT. Demonstration that SV challenges are unhkey is accomplished by analysis results indicating that peak steam line pressure remains a minimum of 25 psi below the SV setpoint.
As set forth in NUREG 0737, Section 11 K3.16, licensees are committed
- to minimizing challenges to the SRVs. In Section 3.4, an evaluation of the impact of the two changes discussed in Section 2.1 is provided to demonstrate no significant increase in SRV challenges.
2.2.4 LOCA Limits The various acceptance criteria that must be met are 10CFR50.46 criteria for emergency core cooling systems, containment design basis, containment heat removal, radiological releases, and LOCA induced structural loads.
The LOCA ECCS analysis demonstrates compliance with 10CFR50.46 requirements. These requirements include:
- Peak cladding temperatures below 2200*F,
- Total cladding oxidation below 17% at peak location,
- Ilydrogen generated in the core below 1%, and,
- Core retaining a coolable geometry.
The LOCA analysis will emphasize small to intermediate breaks in the recirculation loops which cause the SRVs to open as well as steam line breaks outside containment. Results are discussed in Section 4.0. In the LOCA analysis, inoperability of I SRV was not considered since there were no cases for which all four SRVs were challenged. Other LOCA design basis criteria and compliance with these criteria are also discussed in Section 4.0.
2.2.5 Mechanical Loads The alTected structures, systems, and components (SSCs) within the scope of work discussed in this report have been previously analyzed as part of the Mark 1 Containment Long Term Program". The analyses were performed in accordance with the ASME 13 oiler & Pressure Vessel Code, Section 111, Division I, with addenda through Summer 1977. Acceptance criteria for all analyses conform to the applicable requirements developed in support of the Mark I Program W.
All analyses performed in support of this initiative to revise SRV setpoint pressures to a common value and increase the setpoint tolerance were performed in accordance with the methods and acceptance criteria as discussed above. Results are discussed in Section 5.0.
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O 2.3 Computer code, tried 2.3.1 L{on-LOCA NSSS Transient Performance All Non-LOCA NSSS transient response analyses are perfonned with the RETRANM computer code. Thu code has received generic appmval fnun the Nuclear Regulaton Ccunmission (NRC) for application to non-LOCA transien.s. Current YAEC BWR analysis methods with RETRAN, including the application of one-dimensional kinetics, are documented in Reference 8 and approved by the NRC in Reference 9.
2.3.2 LOCA Analysis The LOCA methods consist of the RELAP5YAu" (BWR version) thermal-hydraulic analysis method and the FROSSTEY-2(") fuel performance analysis method The NRC approved the LOCA analysis methods for performing evaluation model licensing application covering the entire spectrum of LOCAs in References 12 and 13.
2.3.3 Mechanical Loads Analysis lhe STARDYNE computer code was used in previous analyses for structures, systems, and components (SSCs) which are afTected within the scope of work discussed in this report. This code was fully verified and industry accepted for the applications used. Analyses perfonned included dynamic response for time vaging loads as well as static and thermal load cases. Computer analysis performed within the scope of this report also utilized the STARDYNE code.
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FIGURE 2-1 OVERPRESSTJRE AND RF ACTOR PROTECTION SYSTEM SETPOINTS WITH 3% TOLERANCES h
1275 l
RPV design pressure,1250 l
1225 .
] T.S. SV setpoint,1240 (2) 1200 -- Min. SV setpoint, 1202.8 (2)
{
l I 1175 -- T.S. ATWS recirc pump trip,1150 3 1150
~55 f ,
Max. SRV setpoints, 11.12.4 (1),1122.7 (2),1133 (1) i O.
1 71125 I
, 5::
i T.S. SRV setpoints, 1080 (1), 1090 (2), 1100 (1) d 1100 l O_
/I l l
Min. SRV setpoints, 1047.6 (1),1057.3 (2),1067 (1)
E 1075 cu O
- 631050 l y T.S. high reactor pressure trip,1055 i
O_
1% 1025 --
1000 --
Normal operating range, 920 - 1050 975 --
. 950 -
925 -- ;
l 900 Page 5 i
O 3.0 NON-LOCA TRANSIENT ANAI,YSIS In this section, non-LOCA transient analysis is presented. The analysis supports the following three changes:
- 2) operation at full rated power with an inoperable SRV.
'lhe analysis is perfonned with approved methodology as discussed in Section 2.3.1. This section of the report is organized as follows:
- In Section 3.1, the features of the plant computer model which had to be modified to perform the analysis are described.
- In Section 3.2, analysis is presented which demonstrates continuing compliance with the ASME overpressure limit (110% of design).
- In Section 3.3, analysis is presented which demonstrates continuing compliance with fuel integrity limits by assuring MCPR 2 the safety limit value . !
- In Section 3.4, analysis and discussion is provided which demonstrate SV challenges are unlikely during an AOT l and no significant increase in SRV challenges.
3.1 SRV & SV Modeline Chances The first step in evaluating the impact of the changes is to modify the representation of the SRVs and SVs in the base RETRAN corewide model used in the current safety analysis. Only the model features important to this analysis are discussed 3.1.1 Chances to Evaluate 3% Setpoint Tolerance Plant model changes are necessary to accommodate analyzing the impact of increasing the as-found SRV & SV setpoint l I
tolerance to *3%.
The peak transient overpressure is maximized when SRV & SV opening setpoints are high. Also, higher transient pressures will cause greater core void collapse and yield more positive reactivity feedback from moderator density. It is conservative to assume high SRV and SV setpoints. All new transient analyses presented in this report will include SRV & SV operating characteristics associated with operation at the upper limit of the *3% setpoint tolerance, except the analysis for SV challenges described in Section 3.1.3.
3.12 Change to Evaluate An Inonerable SRV A goal of this calculation is to support power operation with one inoperable SRV. An inoperable SRV is assumed to be unable to perform its safety relief function within 3% tolerance ofits setpoint. An additional change to the set of changes described in Section 3.1.1 above is developed here to evaluate this mode of power operation. An inoperable SRV is simulated by specifying an arbitrarily high opening setpoint (1x10' psia) for one of the SRVs. With this arbitranly high setpoint, the valve will fail to open during an overpressure transient.
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4 3.1.3 Chances to Evaluate SV Challences Margin to SV lift is analyzed with the limiting AOT for overpressure, the GLRWOBP. This event results in the greatest overpressure for AOTs and should not be confused with the MSIV closure event without direct scram on MSIV position switch. The MSIVC is analyzed specifically for demonstration of ASME overpressure compliance. For evaluating SV challenges the analysis is typically performed with best estimate assumptions to demonstrate that a minimum of 25 psi margin exists to SV lin. The plant model is changed to reflect the expected tolerances of the SRVs and SVs. As found testing has demonstrated the expected tolerances of the SVs and SRVs to be less than 1%. For purposes of demonstration of no SV liR with an inoperable SV, a +1% tolerance is applied to the SRVs and a
-l% tolerance to the SVs. This application ofexpected tolerance and use of the remainder of the assumptions for the ASME code and MCPR compliance analysis provides a conservative prediction of the margin to SV liR.
3.2 Compliance With The ASME Osernrenure Limit In this sectkm, the reactor system response to design basis AOTs is discussed. The two transients analyzed are the MSIVC and the GLRWOBP. These are the two most limiting overpressure AOTs. The results for the GLRWOBP also provide boundary conditions for the hot channel (MCPR) analysis discussed in Section 3.3.
3.2.1 MSIVC Transients The MSIVC transient is examined with RETRAN to demonstrate continuing compliance with the ASME overpressure limit discussed in Section 2 of this report. The limiting case for the current cycle is an MSIVC @ EOFPL with 67B scram times l
without credit for reactor trip on MSIV position sensing. Four cases were analyzed:
8 The first case is the limiting MSIVC for the current plant design. Results are shown in Table 3-1 as the
" Reference" case. ;
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8 The second case is the limiting MSIVC with 3% SRV and SV tolerance. l l
9 The third case is also the limiting MSIVC but includes both changes, i e., 3% SRV and SV tolerance, and an inoperable SRV.
e The fburth case is the limiting MSIVC with an inoperable SRV, and 15% SRV and SV setpoint tolerance. This case provides the limiting setpoint tolerance which yields no margin to the ASME overpressure limit.
Results for these cases are summarized in Table 3-1. Each case dernonstrates that the ASME overpressure limit is met.
Each case also produces a similar transient as typified by results for the limiting MSIVC with both changes (the third case).
The sequence ofevents and transient parameters for this case are shown in Table 3-2 and Figure 3-1. With the SVs assumed to have an opening setpoint at the high end of the 3% tolerance, they do not open during the transients. Although the SVs would be challenged at their Technical Specification setpoint of 1240 psig, these results show they are not required to actuate in order to meet the ASME overpressure limit. Future SRV & SV testing may yield as-found setpoints above the 3%
setpoint tolerance limit. Results for the fomth case with 15% tolerance may be used to support regulatory reporting requirements for this contingency.
3.2.2 GLRWOB Transients lhe GLRWOBP @ EOFPL is also examined with RETRAN to demonstrate continued comphance with the ASME limit.
Aner the MSIVC, it is the second most limiting overpressure transient. It is also the typical overpressure transient which yields the most limiting ACPR. Four cases were analyzed:
S The first case is the limiting GLRWOBP for the current plant design. Results are shown in Table 3-1 as the
" Reference" case.
Page 7
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e The second case is the limiting GIRWOBP with 3% SRV and SV tolerance.
e The third case is also the limiting GIRWOBP but includes both changes, i e.,3% SRV and SV tolerance, and an inoperable SRV. His case pmvides the transient core boundary conditions for the hot channel analysis discussed in Section 3.3.
e ne fourth case is a GIRWOBP, at licensed power plus calormetric uncertainties, with an inoperable SRV, +1%
SRV tolerance, and -1% SV tolerance to demonstrate no SV lift Results for these cases are summarized in Table 3 1. Each case demonstrates that the ASME overpressure limit is met.
Each case also produces a similar transient as typified by results for the limiting GLRWOBP with both changes (the third case). The sequence ofevents and transient parameters for this case are shown in Table 3-2 and Figure 3-2. With the SVs assumed to have an opening setpoint at the high end of the 3% tolerance, they do not open during the transients. The SVs l would not be challenged at their Technical Specification setpoint of 1240 psig and these results confirm they are not required to actuate in order to meet the ASME overpressure limit. These cases am bounded by the MSIVC cases previously discussed in Section 3.2.1. He fourth case, the GLRWOBP, with 1% SRV and SV tolerances resulted in greater than 25 psi margin to SV hit 3.3 Operat8ne MCPR Limits The cunnit hot channel analysis was reviewed and the following information extracted. The transients which generate the most limiting drop in entical power ratio (ACPR) include the Turbine Trip Without Bypass (TTWOBP), the Generator Load Rejection Without Bypass (GI.RWOBP), and .he less ofFeedWater Heater (LOFWII). The LOFWH event does not involve RV pressurization. The TTWOBP and the GLRWOBP are pressurization transients in which the SRVs are challenged.
Earlier SRV operation during these events by a lower opening setpoint would reduce core power through moderator reactivity feedback. The resulting MCPR would be less limiting.
Delayed SRV operation during these events by a higher opening setpoint would increase core power through moderator reactivity feedback. The limiting ACPR occurs in the GLRWOB transient. A hot channel analysis was performed with the com boundary conditions from the GLRWOBP @ EOFPL with 678 scram times and both changes (see case 3 in Section j 3.2.2). Results show that the combined efTects of a 3% setpoint tolerance increase and an inoperable SRV cause the ACPR to increase by 0.02. This is caused by positive moderator feedback from higher pressure in the top part of the core as control rods are inserted. Continuing comphance with fuel integrity limits is obtained when either or both changes are implemented by appropriate changes to the various operating MCPR limits identified in the Core Operating Limits Report (COLR).
3.4 SRV and SV Challences 3.4.1 SRV Challenges As set forth in NUREG 0737, Section IIX3.16, licensees are committed to minimizing challenges to the SRVs. In this section, each of the changes discussed in Section 3.1 is evaluated with respect to a potential increase in SRV challenges.
Current practice regarding SRV and SV setpoints is to assure *1% tolerance is met as required by the ASME Boiler &
Pressure Vessel Cale. As-left setpoints always meet the 1% tolerance. This report demonstrates that as-found setpoints within *3% are acceptable and supported by the safety analysis. Ilowever, as-lef t setpoints will continue to meet *1%
tolerance as required by the ASME Boiler and Pressure Vessel Code. Thus, the probability of a SRV actuation by virtue of a low SRV setpoint is not increased by as-found setpoint tolerance increase from 1% to *3%.
The change to allow full power operation with an inoperable SRV has no significant impact on the probability of SRV actuation.
In conclusion, both changes meet the commitment to NUREG 0737, Section llK3.16 regarding minimizing challenges to Page 8
1 1
o SRVs 3.42 SV Challengqs Plant overpressure response is analyzed to assure that SV challenges are not hkely to occur durmg AOTs. For this purpose, a GLRWOBp identified as the limiting AOT for overpressure was analyzed (see Section 2.2.3). The analysis is typically perfonned with best estimate assumptions to demonstrate that a mmimum of 25 psi margin exists to SV lift for the most linuting AOT. The analysis was conservatively performed assummg:
e licensed power level plus calormetric uncertainties e MST Scram Times e an inoprable SRV e +1% SRV setpoint tolerance e -l% SV setpoint tolerance Results for this case are summarized in Table 3-1. The case demonstrates that the SVs are not challenged at their Technical Specification setpoint of 1240 psig minus a 1% tolerance. The results show greater than 25 psi margin to SV lifl. In conclusion, operation at full power with an inoperable SRV will not cause SV challenges during an AOT. The current Technical Specification limit of c 95% rated power with an moperable SRV is not required.
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e TABLE 3-1 CORE WIDE ANALYSIS
SUMMARY
Peak Normalized Peak PressureN (psig)
Transient Description") IIeat Flux Power RVM Steam Line Reference MSIVC; EOFPL; 67B") 1.9994 1.1995 1261 1230 MSIVC; EOFPL; 67B; 3% Tolerance 1.9997 1.1942 1283 1252 MSIVC; EOFPL; 67B; 3% Tolerance; Inoperable SRV 1.9997 1.1942 1316 1290 MSIVC; EOFPL; 67B; 15% Tolerance; Inoperable SRV 1.9997 1.1942 1375 1348 Reference GLRWOBP; EOFPL; 67B") 3.0883 1.2322 1233 1194 GLRWOBP; EOFPL; 67B; 3% Tolerance 3.0883 1.2322 1252 1216 GLRWOBP; EOFPL; 67B; 3% Tolerance; Inoperable 3.0883 1.2322 1269 1237 SRV GLRWOBP; EOFPL; MST; 1% Tolerance; Inoperable 2.6318 1.1813 1233 1194 SRV;"
ASME Overpressure Limit - -
1375 -
(a) The transient cases are described in more detail in the text. The peak values are extracted from RETRAN output.
(b) The ASME overpressure limit is 1375 psig at the lowest elevation of the reactor coolant system (the lower hemisphere of the reactor vessel).
(c) The pressure in the bottom of the reactor vessel is determined by adding 2.87 psi to the lower plenum pressure in the RETRAN model to account for the elevation effect.
(d) These cases were rerun from the current cycle analysis. The results are shown here for reference in evaluating the effects of the two plant changes.
(c) This case demonstrates the recommended 25 psi margin to SV lift at the Technical Specification setpoint of 1240 psig minus 1% tolerance, Page 10
o e
TABLE 3-2 CIIRONOLOGICAL SEOUENCE OF EVENTS i
Transient Case Event Time (Seconds)
MSIVC; EOFPL; 67B; 3% Start MSIV closure; bypass valves fail closed 0.00 Tolerance; Inoperable SRV liigh flux scram setpoint reached 1.45 Control rods begin insertion 1.73 Peak core power occurs 1.88 ATWS Recirc pump trip setpoint reached 2.38 Threes SRVs open 2.67 MSIVs fully closed 3.00 Recire pump field breakers open 3.38 Peak RV pressure occurs 4.80 Peak steam line pressure occurs 5.10 MSIVC; EOFPL; 67B; 3% Stast MSIV closure; bypass valves fail closed 0.00 Tolerance; Inoperable SRV; l
Direct Reactor Trip Reactor scram on 10% MSIV closure setpoint reached 0.71 Control rods begin insertion 0.99 Peak core power occurs 2.04 ATWS Recire pump trip setpoint reached 2.47 Three SRVs open 2.71 MSIVs fully closed 3.00 Recirc pump field breakers tripped 3.47 Peak steamline pressure occurs 3.96 l 4 GLRWOBP; EOFPL; 67B; 3% Start TCV closure; bypass valves fail closed; Rx tripped 0.00 Tolerance; Inoperable SRV TCVs fully closed 0.31 Peak core power occurs 0.80 ATWS Recirc pump trip setpoint reached 1.40 Three SRVs open 1.49 Recire pump field breakers tripped 2.40 Peak steam line pressure occurs 3.10 Page i1
f MSIVC; EOFPL; 67B; 3% TOL;INOP. SRV MSIVC; EOFPL; 67B; 3% TOL; INOP. SRV 618tl 618tl 10F S . 2OFS 6 1,5 e
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Page 14
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l Page 17
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e 4,0 LOCA ANALYSIS His section provides an assessment for LOCA events to verify the acceptability of operation with increased SRV and SV setpoint tolerances of *3% Operation at full power with an inoperable SRV was not addressed since for all LOCA cases only one SRV is challenged. Since only one SRV is challenged, an inoperable SRV would not change the results of the l LOCA analysis. The various acceptance criteria defined in the VY FSAR that must be met are: 10CFR50.46 criteria for emergency are cooling systems; containment design basis; containment heat removal; radiological releases (10CFR100);
and LOCA induced structural loads. ,
i 4.1 Scone of Analvsis he cmTent LOCA licensing analysis"*, performed in conformance with 10CFR50.46 requirements, ensures that the most limiting combination of break size, break kication and single failure have been considered in meeting the 10CFR50.46 acceptance criteria. The analysis shows that the limiting LOCA event is a break in the recirculation loop, with a break area of 0.6 fP, at the pump discharge kication, with loss of one train of DC power as the single failure. The analysis of Reference ,
14 also shows that breaks at other locations, inchiding the steam line, result in much lower peak cladding temperatures (PCT) l' than treaks in the recirculation line. Further, the analysis also shows that for breaks in the recirculation line larger than 0.4 ft 2, the SRVs would not be challenged because the energy removal at the break provides sufficient depressurization to overcome the etrect of MSIV closure. llence, in assessing the impact of changing as-found SRV setpoint tolerances on 10CFR50.46 criteria, only recirculation line breaks below 0.4 fF need to be reassessed.
The Vennont Yankee design basis containment analysis is for a double ended break at the suction of the recirculation pump.
For this large break size, the rate of depressurization is very fast and the SRVs are not challenged. Consequently, the SRV setpoint tolerance has no effect on the design basis containment analysis, and no further evaluation is required. Containment heat removal system design and Equipment Environmental Qualification are also based on the large double ended recirculation line break. The change in as-found SRV setpoint tolerances and operation at full power with an inoperable SRV does not airect large break LOCA events, and hence no further evaluation of these issues is required.
The Vennont Yankee design base accident fbr radioactive material releases and radiological efTects is a complete severance of one main steam line outside the contairunent. For steam line breaks outside the containment, the MSIVs completely close and tenninate radiological dose releases outside the containment, prior to SRVs being challenged. Ilence, the changes to
< the as-lbund SRV setpoint tolerances and operation at full power with an inoperable SRV will not impact steam line breaks from the perspective of radiological releases.
LOCA induced stmetural kuds are detennined from the large double ended recirculation LOCA or steam line breaks inside the containment. SRVs are not challenged for these type of1.OCA events. Consequently, the change in as-found SRV setpoint tolerances and operation at full power u ith an inoperable SRV does not alTect LOCA induced structural loads, and ,
hence no further evaluation of these issues is required. l l
4.2 Analvile Results ,
l Re LOCA analysis that fanns the current licensing basis was used as the base analysis with which to evaluate the changes in as-found SRV setpoint tolerances and operation at full power with an inoperable SRV. He base analysis0 " was perfomniin confonnance with 10CFR50.46 requirements and included a set of conservative initial operating conditions.
In the base analysis, nominal values were used for the SRV setpoints. These were 1080 psig for one SRV,1090 psig for two SRVs and 1100 psig for the remaining SRV. The SVs, which have a nominal setpcint currently at 1240 psig, were not modeled because the SRVs provided suflicient pressure relief to preclude challenging the SVs.
Page 18
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ne current analysis, which consisted of a series of sensitivity studies on the base analysis, examined the impact of l increasing the as-found SRV tolerances to 13%. De resulting serpoints for opening and closing the SRVs are shown !
in Table 4-2, De analysis examined break sizes below 0.4 ft' in the recirculation line, at the pump discharge location, for each case. Two limiting ECCS single failure assumptions were also addressed: (a) failure of an intset loop LPCI injection valve to open sesulting in only two LPCS system available, and (b) failure of one train of DC power supply resulting in one L.PCS system and one LPCI system. No credit was allowed for HPCI availability. These single failure assumptions were identical to those assumed in the base analysis of Reference 14.
2 The analysis evaluated break sizes down to 0.05 f1. The results of the various sensitivities, expressed in terms of ,
peak cladding temperature (PCT), showed very little impact due to the revised as-found SRV setpoint tolerances of 4 i3%. As with the base analysis of Reference 14, all PCT values were well below the acceptance limit of 2200*F l and all c'her 10CFR50.46 criteria continued to be met. The PCT differences between the sensitivity studies and the base case ranged from -28*F to +32*F regardless of which SRV is challenged. As a function of break size, this is shown in Figure 4-2. Break sizes below 0.05 ft were not evaluated because the PCT shows a steady decreasing 2
trend as break sizes decreased from 0.4 ft2 to 0.05 ft . The conclusion of the analysis is that the current LOCA licensing basis will not be impacted by the change te the as-foand SRV setpoint tolerances.
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d TABLE 4-2 SRV PRESSURE SETPOINTS WITII 3% TOLERANCE SRV Identification Nominal Serpoints -3% Tolerance (psig) Setpoints + 3% Tolerance (psig)
Numbers Setpoints (psig)
Open Close") Open Close(')
- 1 1080 1047.6 1016.2 1112.4 1079.0
- 2 and #3 1090 1057.3 1025.6 1122.7 1089.0
- 4 1100 1067.0 1035.0 1133.0 1099.0 (a) Closing setpoints include 3% blowdown.
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4 FIGURE 4-2 PEAK CLAD TEMPERATURE VERSUS BREAK AREA 2200 .
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Page 21
I 5.0 MECilANICAI,LOAps ANAIJsts The purpose of the analysis is to determine the acceptability of as-found SRV setpoint tolerance changes on SRV piping / supports and the effect of increased discharge loads into the Torus. )
i ne method used is to detemune the load / stress increases in each of the affected areas and to compare these increases to the i margins which exist in the current analyses. In cases where the stress margin exceeds the anticipated stress increase, no {
further analysis is regired. In other cases, the original Mark 1 Program analyses were reviewed and conservatism removed )
to calculate lower, more accurate, actual stress values.
The results of this analysis show that the increase of as-found setpoint tolerance is acceptable for all four lines as they are currently installed at Vermont Yankee. This conclusion applies to all SRV piping, supports, and Tonis structure.
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s \
6.0 CONCLUSION
S The following conclusions regarding two changes are derived from the results of analyses documented in presious sections.
The supportmg safety analysis has been performed in a conservative way so that the twc changes discussed below in Sections 6.1 and 6.2 may be implemented either separately or in combination.
. 6,1 SRV and SV Setnoint Tolerance Increase
'lhe as-found SRV and SV setpoint tolerances may be increased from the current 1% to *3%. Safety analyses assuming the increased tolerances verify the following acceptance criteria:
- 1) De ASME overpressure limit of 110% ofdesign is met. This was verified by re-analyses of the two most limiting o verpressure transients, i e. the Main Steam Isolation Valve Closure (MSIVC) and the Generator Load Rejection Without Bypass (GLRWOBP).
- 2) SV challenges will be muumimi dunng the most limiting AOT. This was verified by an analysis of a GLRWOBP.
- 3) The probability of SRV challenges during overpressure events will not be significantly increased as set forth in l NUREO 0737, Section 11 K.3.16.
- 4) Fuel integrity during overpressure transients will remain assured by adjusting the various operating MCPR limits specified in the Core Operating Limits Report (COLR) to reflect the effect of *3% tolerance.
l 1 5) Re-analysis of the limiting Loss of Coolant Accidents confirms that acceptance criteria listed in Section 2.2.4 are l i met.
! I j 6) A mechanical loads anaiy , cfine SRV discharge piping and the Torus confirmed the integrity of these structures, systems, and components during SRV discharge with the increased setpoint tolerance.
j 6.2 Full Power Operati<m With An Inonerable SRV i From Technical Specification 3.6.D.1, reactor power currently must be maintained s 95% of rated power when one of the j four SRVs is inoperable. This limit has been i:pposed to assure no challenge to the SVs in the event of an AOT. It is pmdent to minimize challenges to the SVs which discharge to the drywell.
i l The limiting AOT dimed in Section 6.1,i c. the GLRWOBP, was analyzed assuming an inoperable SRV while operating i at full power. Results show no SV challenge. The 95% power restriction with an inoperable SRV in the Technical Specifications is not required.
4 l Complete support for operation at full power with both changes is provided by performing all the safety analyses discussed
- in Section 6.1 assuming *3% SRV and SV setpoint tolerances (*1% for SV challenges) and an inoperable SRV. Results i l verify all the acceptance criteria previourij discussed in Section 6.1 are met provided the various operating MCPR limits specified in the COLR are adjusted to reflect both changes. A hot channel analysis of the limiting ACPR overpressure l transient, typically the GLRWOBP, confirmed that a 0.02 increase in the operating MCPR limits bounds this effect.
f Dese analysis results allow the conclusion that a 95% power restriction with an inoperable SRV is not necessary. This does j not preclude the prudence of avoiding prolonged operation with an inoperable SRV.
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7.0 REFERENCES
- 1. Docket No. 50-271, " Operating License DPR-28 Technical Specifications and Bases for Vermont Yankee Nuclear Power Station", through Amendment No.141, October 26,1994.
- 2. Final Safety Analysis Report, Vermont Yarlee Nuclear Power Station.
- 3. Letter from L. II. Heider (VYNPC) to D. G. Eisenhut (NRC), " Submittal of Information on NUREG-0737, item 11.K.3.16, ' Reduction of Challenges and Failures of Relief Valves' and item 11.K.3.18, ' ADS Actuation'", FVY-8180, May 15,1981.
- 4. Technical Report TR-5319-1, Revision 2, Mark 1 Containment Program, " Plant Unique Analysis Report of the Torus Suppression Chamber for Vermont Yankee Nuclear Power Plant", November 30,1983.
- 5. Technical Report TR-5319-2, Original Issue, Mark 1 Containment Program, " Plant Unique Analysis Report of the Torus Attached Piping for Vermont Yankee Nuclear Power Plant", October 3,1983.
- 6. NEDO-245831-1, " Mark 1 Containment Program Structural Acceptance Criteria Plant Unique Analysis Application Guide", General Electric, October,1979.
- 7. EPRI NP-1850, Volume 1, Rev. 4, "RETRAN A Program for Transient nermal- Hydraulic Analysis of Complex Fluid Flow Systems, Volume 1: Theory and Numerics (Revision 4)", November,1988.
- 8. YAEC-1693, " Application of One-Dimensional "inetics to Boiling Water Reactor Transient Analysis Methods", June,1989.
- 9. Letter from A. C. Thadani (NRC) to R. W. Capstick (VYNPC), (SER), December 12, 1989.
- 10. R. T. Fernandez, et. al., "RELAPSYA -A Computer Program for Light Water Reactor System Thennal-Hydraulic Anrlysis," YAEC-1300-P-A, October 1982, Revised August 1993.
I 1. Letter from L. W. Capstick (VYNPC) to V. L. Rooney (NRC), " Vermont Yankee LOCA Analysis Method FROSSTEY Fuel Performance Code (FROSSTEY-2)", FVY 87-116, Vermont Yankee Nuclear Power Corporation, December 16, 1987.
- 12. Letter, V. L. Rooney (USNRC) to R. W. Capstick (VYNPC), " Approval of the Use of Thermal-Hydraulic Code RELAP5YA, (TAC No. 60193)," NVY 87-136, August 25,1987. (Docket No. 50-271).
- 13. Letter, P. Sears (USNRC) to L. A. Tremblay (VYNPC), " Safety Evaluation for Vermont Yankee Nuclear Power Station RELAPSYA LOCA Analysis Methodology"(TAC No. M74595), October 21,1992. (Docket No. 50-271).
- 14. YAEC-1772, " Vermont Yankee Loss of Coolant Accident Analysis," L. Schor and F. Seiface, June 1993.
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