ML20199G155
| ML20199G155 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 11/18/1997 |
| From: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Collins S NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML19313D093 | List: |
| References | |
| CAW-97-1188, NUDOCS 9711250108 | |
| Download: ML20199G155 (57) | |
Text
~ :
m.
_.m.
. ~.. -.. -.-.-_ _
d x
~
f g g ggg-Nuclear Sunes Division :
6:
- Electric Corpofstion wm._.
1
=
Pmsburgh Pennsylvania 15230-0355 I
I
- November 18,1997^
- CAW-971188:-
T Document Control Desk
- U.S. _ Nuclear Regulatory Commission Washington, DC 20555:
= Attention: - Mr. Samuel J. Collins J
- APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE
Subject:
SNC Response to NRC RAI on BELOCA," NSD sal ESI-97-646 (Proprietary) :
Dear' Mr. Collins:
The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-97-l!88 signed by the owner of the proprietary information,
- Westinghouse Electric Corporation. The affidavit, which accompanhthis letter, sets forth the basis on which the information may be withheld from public disclosure b, ne Commission and addresses wit!t specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.790 of the Commission's regulations, Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Southern Nuclear Operating Company.
Correspondence with respect to the proprietary aspects of the application for withholding or the 1 Westinghouse affidavit should reference this letter, CAW-97-Il88, and should be addressed to the undersigned.
Very truly yours, N. J. Liparu o, Manager Equipment Design and Regulatory Engineering-Enciosures cc: Kevin Bohrer/NRC (12H5) -
g f
PDR m
. m e-n gq,,.
-1 k-6-
~um
.o u
-m e
-.4-g 4
...,r,en-
,.n',-
'--+,w
~,-w-,k-o-
w
CAW-97-Il88 -
AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:
sS COUNTY OF ALLEGHENY:
s Before me, the undersigned authority, personally appeared Henry A. Sepp, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Aflidavit on behalf of Westinghouse Electric Corporation (" Westinghouse") and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:
L6 Lud Henry A. Sepp, Manager l H.'
i Regulatory and Licensing Engineering Sworn to and subscribed before me this / [
day of h b u q [ d,1997 Notana! Seal Janet A Schwab. Notary PutAc MC ni a t.
re May 22 2 Member,PentsyNama Assocation of Notanes
/
Notary Public 4
. CAW-97-1188 (1)
I am Manager, Regulatory and Licensing Engineering, in the Nuclear Services Division, of the Westinghouse Electric Corpension and as such, I have been specifically delegated the function of reviewing the proprietary intormation sought to be withheld from public disclosure in connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Energy Systems Business Unit.
(2)
I am making this Affidavit in conformance with the provisions of 10CFR Section 2.790 of the Commission's regulations and in conjunction with the Westinghouse application for withholding accompanying this Affidavit.
(3)
I have personal knowledge of the crheria and procedures utilized by the Westinghouse Energy Systems Business Unit in designating information as a trade secret, privileged or as confidential commercial or financial information.
(4)
Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Commission's regulations, the following is furnished for coasideration by the Com. mission in determining whether the information sought to be withheld from public disclosure should be withheld.
(i)
The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.
(ii)
The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required, i
I Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:
1 1
237 TORI %2.lllN7
~
~-
-3 CAW-97-1188 (a)
The information reveals the distinguishing aspects of a process (or component,
- structure, tool, method, etc.) where prevention ofits use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.
(b)
It consists of supporting data,-including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures _
a competitive economic advantage, e.g., by optimization or improved marketability.
(c)
Its use by a competitor would reduce his expenditure of resources er improve his competitive position in the design, manufacture, shipment, insta"ation,_
assurance of quality, or licensing a similar product.
(d)
It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.
2 (e)
It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.
(f)
It contains patentable ideas, for which patent protection may be desirable.
There are sound policy reasons behind the Westinghouse system which include the following: -
(a)
The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.
(b)
It is information which is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.
mcenim l
i r
-4 CAW-971188 (c)
Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.
(d)
Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.
(e)
Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.
(f)
The Westinghouse capacity to invest corporate assets in research and development dependa upon the success in obtaining and maintaining a competitive advantage.
(iii)
The information is being transmitted to the Commission in confidence and, under the provisions of 10CFR Section 2.790, it is to be received in confidence by the Commission.
(iv)
The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and bdief.
(v)
The proprietary information sought to be withheld in this submittal is that which is appropriately marked in "SNC Resp (mse to NRC RAI on BELOCA,"
NSD-sal-ESI-97-646 (Proprietary), November,1997 for Farley Units I and 2, being transmitted by Southern Nuclear Operating Company letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk, Attention Samuel J. Collins. The proprietary information as submitted for use by Southern Nuclear Operating Company for Farley Units I and 2 is expected to be muunnw
4
-5 CAW-97 Il88 applicable in other licensee submittals in response to certain NRC requirements for justification of the use of the Best Estimate LOCA methodology.
This information is pan of that which will enable Westinghouse to:
(a)
Provide Best Estimate LOCA methodology.
(b)
Provide Farley specific application of Best Estimate LOCA methodology.
(c)
Assist the customer in obtaining NRC approval.
Further this information has substantial commercial value as follows:
(a)
Westinghouse plans to sell the use of similar information to its customers for purposes of meeting requirements for licensing documentation.
(b)
Westinghouse can sell support and defense of the technology to its customers in the licensing process.
Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar methodologies and licensing defense services for commercial power reactors without commensurate expensea. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.
The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money, in order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, womunm
6--
CAW 971188 having the requisite talent and experience, would have to be expended for developing.
e
- - testing and analytical methods and performing testing.'
Further the deponent sayeth not.
I 1
1
+
1 6
L
+
L N 'kMIkklN
e
.e L
Proprietary Information Not%e -
- Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC In connection widi requests for generic and/or plant-specific review and approval.
~In order to conform to the requirements of 10 CFR 2 '90.of the Commisslor.'s regulations concerning the protection of proprietary information so submitted to the NRC, the information'which is proprietary o'
in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only_ the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) contained within parentheses located as a superscript immediately following the brackets enclosing each item of information being identified as proprbtary or in the margin opposite such
. information. These lower case letters refer to the types of information Westinghone customarily holds in confidence identified in Sections (4)(li)(a) through (4)(ll)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.790(b)(1).
e
t l
Copyright Notice
+
The reports transmitted herewith each bear a Westinghouse copyright notice. He NRC is permitted 'to make the number of ceples of the information contained in these reports which are necessary for its internal use in connection with generic and plam-specific reviews and approvals as well as the issuance,
+
denial, amendment, transfer, renewal, modification, suspension, revocation, or.jolation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.790 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be reqaired by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary natice if the original was identified as proprietary.
t 6
6
(
f
,5 l
ATTACilMENT I ~
.SNC Response to NRC Request For Additional Information
~
Related To Power Uprate Submittal-Joseph M. Farley Nuclear Plant, Units l & 2 SNC RESPONSES TO NRC QUESTION NOS. 1 - 11 & 13 -33
+
d
+
t v
+
w re-
= - - - + - *
- + - - - - - - - * - = -
-~'-
r+-
SNC Response To NRC Request For Additional Information Related To Power Uprate Submittal-Joseph M. Farley Nuclear Plant Units 1 & 2 NRC Ouc1 tion No.1 The response to question 1 in the August 5,1997, submittal does not completely answer the
. question; The question asked for a Standard Review Plan (SRP) 5.2.2 analysis to be performed at the new power level. He SRP 5.2.2 analysis differs from a Chapter 15 FSAR type analysis q
because SRP 5.2 2 does not allow credit for the first safety grade reactor trip. Additionally, the l
SRP 5.2.2 analysis that was performed at initial licensing, and referenced in the response to the question, no longer applies because the reactor power will be increased. Please provide an SRP 5.2.2 analysis to show that the primary and secondary overpressure protection is adequate.
Indicate what percentage of the relieving capacity is used to mitigate the event and the peak primary and secondary pressure.
SNC.lkspanstEod Section 5.2 of the Farley Nuclear Plant (FNP) FSAR references WCAP-7769 Revision 1,
" Overpressure Protection for. Westinghouse Pressurized Water Reactors," June 1972. This, Overpressure Protection Report (OPPR) was used to establish the primary and secondary system safety relief valve sizing requirements in support of the plant pursuing the operating licensing, as documented in the initial edition of the Farley FSAR.
WCAP-7769 Revision 1 (page 2-1) specifically notes that the pressurizer and steam generator safety valve sizing calculations assume an initial condition corresponding to the Engineered Safeguards Design Rating (ESDR). With respect to the values listed in WCAP-7769 Revision i Table 2-1 (page 2-4), the uprated power level and valve sizing of the FNP are most similar to the Beaver Valley Power Station reference configuration. (FNP and Beaver Valley Power Station are both 3-loop Westinghouse PWR's of similar design and vintage.)
Comparison of Beaver Valley Power Station Reference Configuration to Uprated Farley Nuclear Plant Parameters Beaver Valley-Uprated Farley htra! peter Power Station Nuclear Plajn ESDR Core Power, MWt 2774 2775 No. of PRZ Safety Valves 3
3 Total PSV Capacity, Ibm /hr 1,035,000 1,035,000 No. of SG Safety Valves 15 15 Total SGSV Capacity, 12,148,647 12,880,266*
lbm/hr
- Minimum total relief capacity assumed in the FSAR Loss of External Load and/or Turbine Trip analysis (Section 15.2.7).
Al - 1 L:
Although there exists an insignificant difTerence in the assumed initial power level, the Farley safety relief valve capacities meet or exceed the reference values. Therefore, the valve sizing basis rentsns valid for the proposed FNP uprate, and the SRP 5.2.2 analysis, which was performed to establish sizing requirements, has not been revisited.
Regardicss of meeting the WCAP-7769 Revision 1 overpressure protection basis, the reference analysis following Standard Review Plan Section 5.2.2, and supporting this WCAP, was applied for FNP in support of obtaining the initial operating license. For subsequent licensing amendments, the limiting FSAR Chapter 15 safety analysis, i.e., Loss of External Load and/or Turbme Trip (LOl/IT) analysis presented in Section 15.2.7, has provided the basis for meeting the overpressure protection requirements of the ASME Boiler and Pressure Vessel Code. This analysis demonstrates that the overpressure protection, provided by the Reactor Trip System and pressurizer and steam generator safety ulves, i:: adequate to prevent the primary and secondary system pressures from exceeding i 10?' af their respective design pressures.
The LO1/IT analysis perfonned for the FNP uprate was performed using the same methods as the analysis currently presented in the FNP FSAR (Section 15.2.7). The analysis method includes the guidelines documented in WCAP-12916-P-A," Pressurizer Safety Valve Set Pressure Shifl,"
which was approved in February 1993. The LOl/IT analysis continues to meet the 110% of system design pressure criterion. The results of this analysis were presented in WCAP 14723 (page 6-82) and are briefly discussed in response to Question No. 26 herein. The peak RCS pressure (at *.he RCP outlet) calculated in the analysis was 2747 psia compared to a conservative limit value of 2748.5 psia. The peak main steam system pressure calculated in the Chapter 15 analysis was i 199 psia compared to a limit value of 1208.5 psia.
W/gls 11/12N7 N.RC_0mslinaNo_2 The response to question 4 does not completely answer the question. The response indicates that only a review of the FS AR was performed to identify areas for review. 'lo determine the risk to public health and safety, please perfonn a review of your Individual Plant Examination and determine if any of the compensatory actions or success paths identi6ed are no longer available at the uprated conditions (power, reactor coolant system (RCS) flow, emergency core cooling system (ECCS) flow).
SNC Response No 2 The original Farley IPE model is no longer maintained by FNP. A plant specine PRA model using the EPRI CAFTA sonware is now employed. The Farley PRA uses a combination of FSAR and PRA-specific success criteria. Where the FSAR is used as the basis of PRA success criteria, it has been verified that the FS AR criteria following power uprate are consistent with the PRA assumptions. The PRA-specific analyses were performed using a combination of the Modular Accident Analysis Program (M AAP) 3.0b code and the Westinghouse Transient Real-time Engineering Analysis Tool (TREAT) with Farley-specific models Due to the unavailability of the Farley-specific TREAT input files used for the success criteria development, all success criteria and operator action timing analyses used in the updated Farley PRA were re-analyzed at the power Al-2
,I uprate conditions using a Farley-specific NAAP 4.0.3 model. De results indicate that the power uprate being requested will not have any impact on the success criteria or operator action failure probabihties used in the current Farley PRA model. Therefore, there will be no impact on the calculated Core Damage Frequency or Large Early Release Frequency due to the power uprate.
This includes modeling the effects of all uprate changes such as RCS core flow, ECCS injection flowrates, uprated power, etc.
SNC/dem & rah 11/17/97 NRC Ouestion No. 3 The response to question 5 indicates that "the Farley Units are currently operating with ziiealoy clad fuel and ZlRLO clad fuel (VANTAGE +)." If both VANTAGE 5 and VANTAGE + fuels will be used, please update both the appropriate sections of the TS and FSAR with appropriate references to the VANTAGE + fuel. The safety evaluation referenced in the response to question 5 indicated that the NRC approved the use of ZIRLO clad for lead test assemblics. Pkase verify that all the analyses presented in WCAP-12610-P A," VANTAGE + Fuel Assembly Reference Core Report" have been performed with eccep'.able results.
SNC Renmdo_2 Both VANTAGE 5 and VANT. IGE + fuel designs are utilized in the Farley units. VANTAGE + is a fuel-type nomenclature used for VANTAGE 5 with ZlRLO" cladding. The Techni,al Specifications and FS AR have been updated to account for ZlRLO* clad fuel. All analyses, as specified in WCAP-12610-P-A, have been completed for both Farley units to document and demonstrate the acceptability of ZIRLO* clad fuel. The analyses results meet applicable criteria and are therefore acceptable. SNC submittal dated June 30,1997, Joseph M. Farley Nuclear Plant Technical Specifications Change Request Credit for Boron for Spent Fuel Storage" requested administrative changes to remove references to specific fuel-type nomenclature (e.g.,
LOPAk, VANTAGES, etc.) in the Design Feature section of the Farley lechnical Specifications.
Subsequently, SNC identified explicit use of fuel 4ype nomenclature in Bases 2.1.1, " Reactor Core." Therefore, SNC will provide fuel type nomenclature corrections to the Farley Technical Specifications llases.
W/rgm 1I/14/97 A SNC/mge-It/17/97 NILC_Qtesion No. 4
'I he response to question 7 indka:cs that the Tccimical Specifications (TSs) limit Tm to below 580.3 F; however, the plant is not naalyzed for operation at this temperature. Because the existing TS does not limit Tm to within the accident analysis window, with appropriate instrument uncertainties, provide an accident analysis that uses the appropriate TS Tm window, with instrument uncertainties.
Al - 3
SEC.Essanse No. 4 As dis:vssed in WCAP 14723 Section 6.0.1; six operating parameters, including RCS Tavg, ate explicitly modeled in various transient and accident analyses with initial condition steady-state uricertainties. A 6'F uncertainty allowance was assumed for Tavg For DNB transients using the Revised Thennal Design Procedure (RTDP), the Tavg uncertainty is included in the derivation of the core safety limits. For non-LOCA analyses using the Standard Thermal Design Procedure (STP) and LOCA analyses, the Tavs uncertainty is added to the initial condition steady-state Tavg.
For uprate, the upper temperature limit of the Farley accident analyses is derived by adding a 6.0*F uncertainty allowance to the upper operating limit of the full power Tavg window of 577.2 F (i.e.,
577.2 + 6 = 583.2'F). A Farley specific Tavg indication uncertainty of 2.9'F has been included in the Technical Specifications DNB Tavg limit. The resulting surveillance limit for indicated Tavg (DNB parameters, Table 3.2 1) is 583.2 - 2.9 = 580.3*F.
The plant control systems are set and calibrated for the maximum target temperature (or less),
where the RCS Median Tavg = NSSS Control System Tref = 577.2*F. The plant operator periodically verifies proper control system operation using the RCS Tavg indicators. As long as the indicated average Tavg is s 580.3 F during steady-state full power conditions, the unit is operating within the bounds of the accident analyses. Therefore, the proposed Technical Specifications change to the DNB indicated Tavg limit from 580.7'F to 580.3'F is appropriate.
W/whm - 1IIl4/97 & SNC/mge - l1/17/97.
NRC Onestion No. 5 The response to question 10 does not provide sufficient detail regarding the analysis of the 50%
load rejection. Please describe in greater detail the calculated results including the magnitude (requested in question 10 and not provided) of the power and temperature oscillations. Evaluate the ability of the analysis techniques to adequately evaluate the core transient. Additionally, is there any other operating window where greater oscillations can occur?
SNC_Ecsponse No 5 The 50% load rejection analysis was performed using the LOFTRAN computer code. This code has been used for over 20 years by Westinghouse in the analysis of certain PWR Condition I,11 and III events. The results of the LOFTRAN analysis compare favorably with the actual plant behavior during operating transients. Generally, if all control systems operate as expected and as analyzed, LOITRAN predicts a conservative margin to the primary side protection system setpoints. In addition, LOFTRAN is a conservative predictor of the potential for plant oscillatory behavior (i.e., if LOFTRAN predicts an oscillatory response, plant response is more stable than the LOFTRAN prediction).
The intent of the 50% load rejection analysis as discussed on page 4-20 of the NSSS licensing
- report was to determine the operating margins and control system stability rather than the core response. Although, the LOITRAN code has the capability to determine the core response (i.e.,
Al - 4
DND, core subcoolint, etc.) for a 50% load rejection transient, the core response was not performed because this is a Condition I transient. An equivalent (i.e., bounding) core response transient has been addressed in the FSAR Chapter 15 Condition 11 Loss of Loadn'urbine Trip analysis.
He nature of the oscillatory plant response following a 50% load rejection transient from 102%
full power at low Tavg (567.2 F) showed a decaying (damping) oscillatory response in both primary (nuclear power, Tavg, pressurizer pressure) and secondary (steam flow, steam pressure and temperature) parameters with five oscillations being noted before plant stability is reached.
%e nature of the peak oscillations for various parameters were as follows.
Nuclear power exhibits a decreasing trend throughout the transient due to automatic rod control action. The maximum and minimum nuclear power were 102% of full power (initial value) and 52% of full power respectively. Although the nuclear power change is a decreasing trend throughom the transient, a increase in nuclear power was noted at approximately 65,125,200,260 and 350 seconds in the transient. Rese five peaks in nuclear power occur at approximately 80%,
72%,68%,63% and 59% of full power. The nuclear power increases by approximately 1% above these values (maximum) in about 20 seconds and then starts to drop.
The pressurizer pressure shows approximately 1100 psi peak to peak variations, with pressure not exceeding 2330 psia or dropping below 2040 psia.
The Tavg shows approximately i 3.5*F peak to peak variations with the peak Tavg being 6*F above the initial Tavg of 567.2'F.
The steam pressure show, approximately i 45 psi peak to peak variations, with pressure not exceeding 850 psia or dropping below the initial value of 680 psia.
The oscillatory behavior is worst at the very beginning of core life (as it was assumed in the analysis). As the core burnup increases, the moderator temperature coefficient (MTC) becomes more negative which will result in a more stable (less oscillatory) plant response. As stated in the NSSS licensing report, the 50% load rejection analysis performed at low Tavg with end of core life conditions showed stable plant response in addition, Tavg of 567.2 F is the lower limit on the full power Tavg operating window. A full power Tavg value above this will result in a more negative MTC and less oscillatory response. Therefore, the analyzed condition is considered the worst-case operating point, and there is no other operating temperature within the analyzed window which will result in greater oscillatory response.
- Wisma. II!!197 NRC Ouestinn No. 6 Please provide a summary of calculations that show that the atmospheric dump valves are sized adequately to maintain a 50 F/hr cooldown rate over the entire range that they are necessary using the higher decay heat associated with the power uprate. The response to question 9, the submittal description, and the FSAR description do not provide sufficient detail.
AI - 5
SNC Response No. 6 he capacity of the steam generator atmospheric relief valves is sufficient to achieve a cooldown rate of 50*F/hr throughout the entire cooldown to permit the cooldown to be completed with the normal auxiliary feedwater system water supply. The cooldown npe ation is completed when the reactor coolant system hot leg temperature tias reached 350 F, which permits alignment of the residual heat removal system. De 350*F residual heat removal cut-in point con esponds to the lowest steam generator steam pressure, and since atmospheric r: lief valve capac ty is a function of
' inlet pressure, this point in the cooldown is limisg with respect to valve sizing. Therefore valve capacity at the uprated power icvel was va ed at the end of cooldown; i.e., abo ut 6 hrs idler reactor trip assuming 2 hrs at het candby and a 50 F/hr cooldown rate.
The uprate evaluadan confirmed atmospheric relief valves capacity at the end of cooldown by converting the maximum steam dump capacity of the valves at this point in time to a heat removal rate by multiplying the capacity of the valves in lbs/hr by the difference between the enthalpy of the auxiiiary feedwater and the enthalpy of steam at the loviest steam generator pressure. The maximum coutdown rate was then determined by first subtracting from the atmospheric relief valves heat removal capacity the reactor decay heat generation rate m the end of the cooldown period (about 6 hrs) and then dividing the remaining steam dump heat removal capacity (in Btu /ht) by the overall reactor coolant system heat capacity (in Btu / F). This analysis indicated, at the Farley uprated power level, the maximum achievable cooldown rate exceeded the required rate and therefore verified that the steam generator atmospheric relief valves have more than sufficient capacity to achieve a design bases cooldown at the uprated conditions.
W/rgm. I1/14/97 NI!C Ouestion No. 7 Arc a!! codes used in the power uprate and the current licensing basis up-to-date with respect to relevant changes in knowledge, regulations, guidance, and changes in the plant? Discuss this in detail for all codes. Are the codes used in confom ance with their limitations and restrictions?
ENC Response No. 7 in response to this question, Farley designers and cognizant engineers were asked for input. The SNC position for the Farley uprate licensing submittal is that the principal codes used in support of the submittal are up-tostate with respect to applicable codes and standards and have bcen used in conformance with their limitations and restrictions. A listing of the principal codes used in the power uprate and the current licensing basis is provided in the SNC response to Question No. 8 below.
With regard to plant design changes, the effects on current analyses are reviewed as part of the 10 CFR 50.59 process prior to perfonning those changes. Furthermore, the joint SNCAVesiinghouse cycle-specific reload process explicitly identifies all planned / scheduled plant changes :md independently assesses the potential impact on the safety analyses for each reload core design.
AI - 6
For the uprate project a comprehensive review of plant configuration was performed by a team, rhich included the FNP designers and key FNP site and corporate support personnel. Applicable changes were incorporated in the Uprate Analysis input Assumption List to ensure that uprate analyses and associated assumptions actually reflect plant as built configuration as well as plant changes and operation-at uprated conditions.
W/rgm & SNC/mjc & ruge - 11/17/97 NRC Ouestion No. 8 Identify all codes used for this power uprate, and of these, which were previously seviewed or appravcd for use at your plant. To be able to detennine the effect on the safety margin from u.ing the new code, if a new code is used, what would a baseline run at the 01:1 power level reveal in tenus of calcula'ed results (i.e., compare the peak cladding temperature (PCT) at the new power relative to the PCT at the old power)? Besi des this baseline run, identify all cumulative decreases in safety margin from successive plant and procedure modifications. Provide a justification for the new code's applicability to your plant.
SNC Rummd9J!
Consistent with the Farley FSAR (Section 15.1.9), the following tablu provides a list of the principal codes used for Chapter 15 analyses. The table also delineates which codes have been generically approved by the NRC and which codes have been previously used and approved for Farley.
In response to General Qaestion No. 2 of the previous RAI (SNC letter to NRC dated August 5, 1997), SNC provided a list of new methods used for the Farley power uprate. No quantitative approaches were made to baseline new codes to older codes. For analyses, such as BELOCA, it is virtually impossible to provide a benchmark calculation at the current power level because of the complexity and the sheer volume of the calculation without re-performing a significant set of the calculations. For new principal codes used in the uprate analyses the following is offered.
W/rgm 11/15/97 MULTIFLEX MULTIFLEX 1.0, LATFORC and FORCE 2 were previously used on Farley Unit I for the vessel analysis during the upflow conversion analysis. This is the first application of MULTIFLEX 1.0 to Farley Unit 2, and it replaces the prior BLOWDOWN 2 analym Since steam generetor forces and loop forces were not reanalyzed for the uprate (evaluations ;oncluded the current steam generator and loop forces remained bounding for the uprated conditions), those codes are not changing.
Al - 7
- n. - -,
E j
1 LOCA forces are not substantially impacted by power changes, except as power feeds back into
. changes in RCS initial temperatures and pressurcs. In this case, the p6wer uprate change in RCS '
-j conditions would be conservatively estimated to result in an increase in forces of 13.2% or less.
- liowever, there is no particular acceptance criterion on LOCA forces. -These forces are translated u into system loads using a dynamic non linear modelf As a result, the change _in' applicable system 3
- LOCA_ loads can be substantially greater or less than the change in LOCA forces. It is the system '
4 loads which ultimately translate into acceptance criteria in the calculated consequences with :
respect to structural integrity, f As' indicated in the WCAP 8708 (Reference 1) SERF the LOCA forces benefit of using ;
' MULTIFLEX 1.0 rather than a rigid calculation like BLOWDOWN 2 is estimated to be on the order of 30% reduction in calculated peak loads; Also of note, the LOCA forces analysis has taken credit for NitC approved Leak Before-Break exemptions to the main loop piping.- This results in
- reduced break areas assumed for purposes of calculating LOCA forces. The benc6t of this area reduction is estimated to be approximately 30%. In other words, LOCA forces are estimated to go up by 13.2% or less based on the uprating alone. Offsetting this increase are an ~30% benefit from break area reduction for Units I and 2, and an ~30% benefit from the use of MULTIFLEX l.0 methodology for Unit 2 only. In the net, LOCA forces are expected to decrease as a result of reanalysis, although the benefit in terms of calculated loads is not readily quanti 6able because of the non linearity of the dynamic structural models.
i DELOCA The code used for the Best Estimate Methodology LBLOCA uprate analysis is WCOBRAffRAC Mod 7A, Rey, l.' The Best Estimate Methodology using ECOBRA/ TRAC has received generic USNRC approval for use in the analysis of 3 loop and 4 loop plants with ECCS injection into the cold legs (Reference 2). Farley Units 1 and 2 are 3-loop plants with ECCS injection into the cold legs and are therefore covered by the USNRC SER.
The following table shows the major differences in plant conditions between the non-uprated and e
uprated analyses which affect the large break LOCA. As can be seen from the table, the uprate analysis has included increases in operating margin for the plant. Use of the DE LBLOCA methodology has allowed for these increased operating margins without a signi6 cant increase in -
. peak cladding temperature. Instead, the increases are accomplished with littic change in margin to
~ the 2200 F regulatory limitc 4
J a
Al - 8 4
rr-4 sa@
4
+
+
e r-w.
m 4-
-e
+
+ -, -
e-4-
c -
v-
- r w-- w-
,--,r-t w
v
Comparison of Uprated Conditions to Non-Uprated Conditions for Faricy Units I and 2 LBLOCA Analysis Earamgle.r.
Pre-Up.r_atg_Yahg IJprate Value (BART/ BASH) '
(BELOCA)
NSSS Power (MWt) 2660-2785 TS F, Max 2.45 2.50 Q
TS FAli, Max 1.70 1.70 PbarilA, Max 1,48-1.574 RCS Thermal Design 87,200 86,000 Flow per loop (GPM)
ECCS Flow 8%
10 %
Degradation RCS Vessel AvEge 577.2 567.2-577.2 Temperature ( F)
PCT ( F) 2052
<2064 ORIGEN With regard to radiological analy ses that utilized the original ORIGEN code and the revised ORIGEN2 code to provide source term information, both code versions were/are industry standard codes based on the latest industry experimental d*.a. This data varies isosope-by-isotope.
Therefore, a single summary cannot be provided that would be applicable to all spectra of accidents (e.g., some isotopes go up and some go down). In general the da'a are up-to-date, well documented, and accepted by the industry and are therefore appropriate for the Farley uprate analyses.
TACT
- Radiological consequences of postulated accidents were evalaated using TACTS, TACTS was
- developed for the NRC for use in evaluating radiological cc.isequences and has been validated for safety-related applications by Southem Company Services.
The major model assumptions pr:viously provided in response to General Question No. 2 (SNC letter to NRC dated August 5,1997) are consistent with the current plant configuration and applicable Regulatory Guides as described in the BOP Licensing Report. As noted in the
' introduction to NUREG/CR$ 106, "the mathematical treatment within the code has remained largely unchanged from earlier versions." Thus recent submittals which have been prepared with codes derived from the earlier TACTill would be expected to yield results similar to those obtained with TACT 5. The original FSAR radiological consequences submitted with the request for an Al - 9
l operating license were prepared with a number of computer codes used to evaluate simplified (e g.,
single region containment) release and transport models. No comparison of these codes has been i
made; however, the equations governing transport of the released activity are similar to those evaluated in the mathematical mod;l of TACT 3; hence, no significant code based differences would be expected.
GOTillC The LOCA and MSLB containment pressure / temperature response analyses were performed using the GOTrilC computer code, version 5.0c. GOTillC was developed under EPRI contract from the older NRC code, FATil0MS. GOTillC was developed under a fully qualified quality assurance program and has undergone extensive peer review. GOTlilC has been validated for safety related applicatioru at Southem Company Sersices.
There is little difference in the original / current FSAR model (i.e., COPATTA/ COMPACT) and the GOTillC model used in the uprate containment analyses. The principal difference in the two mialyses is the change in the blowdown data (section 6.4 of NSSS Licensing Report). Other model/ input changes were: a) RiiR heat exchange properties (hcat transfer coeflicient and CCW flow) were modified to represent actual plant design data; b) Containment cooler performance represents highly degraded cooler service conditions (i.e., the model uses a sing!c (degraded) cooler in sersice); c) Initial containment temperature is assumed to be 127 F, which corresponds to the containment design operating bulk average temperature plus margin; and d) Initial containment pressure is assumed within the range of-1.5 to +3 psig, consistent with the Technical Specifications. As noted in the BOP Licensing Report, the uprate analyses made maximum uso of the previous analyses for input assumptions. In addition, the previous blowdown model data were used with the GOTillC modeling and compared to the previous analyses with consistent results, with differences on the order of I psi and 10 F at peak conditions.
Ess,onse No. 8 References 1.
WCAP-8703 P-A,"MULTIFLEX, a FORTRAN IV Computer Program for Analyzing ncrmal-llydraulic Structure System Dynamics," September 1997 (Westinghouse Proprietary).
2.
Letter, R. C. Jones (USNRC) to N. J. Liparuto (W), " Acceptance for Referencing of the Topical Report WCAP 12945 (P), Westinghouse Code Qualification Document for Best Estimate Loss-of-Coolant Analysis," June 28,1996.
f Al - 10
~
~.
?
FARLEY-NUCLEAR PLANT --
= -
- FSAR CHAPTER 15 ANALYSES
SUMMARY
OF INITIAL CONDITIONS AND PRINCIPAL COMPUTER CODES USED NRC-
- ' Initial NSS A
i Computer Approved :
1hermal Power ~
Assumptions Analing Codes Utihred.-
- Genenc/FNP Output Assumed
' thed Uncontrolled RCCA bank 1WINKLF Yei/Yes 0 (suberitical)")
Yes -
withdniwal from a sub-
-FACTRAN,-
cntical condition 1111NC
' Uncontrolled RCCA bank LOF1RAN I Yes/Yes
-279,1674,4 Yes withdrawal at power and 2790"
- RCCA misalignment
'llllNC, ANC, Yes/Yes 277594 Yes -
LOITRAN Uncont:Jted boren dilution.
_NA NA/NA.
O and 2785(a)
Yes Partialloss of forced reactor -
LOITRAN,11tlNC Yes/Yes 2790(d Yes coolant flow -
- FAC1RAN, Startup of an inactive RCP
.NA NA/NA NA NA loss of external diectrical LOITRAN Yes/Yes 2785(*
Yes load and/or turbine trip _
- loss of normal fxdwater.
LOFTRAN Yes/Yes 2785(**
Yes
'lhss of all ac power to LOITRAN Yes/Yes 2735 e>
Yes 9
=
~ the station auxiliaries td Yes Excessive heat removal due to 1.OITRAN Yes/Yes 0 and 2785 i :-
feedanter system malftmetions
+
Excendve land increase LOFIRAN Yes/Yes 2785(d Yes Accidental depressurization LOFfRAN Yes/Yes 2785(d Yes of tic RCS c AcciJental depressurization LOITRAN Yes/Yes 0 (suberitical)(*
Yes of the main steam system inadvertent operation of the LOFIRAN Yes/Yes 2785(uh Yes ECCS dunng power
- operation
- Loss of reactor coolant from
- NOTRUMP, Y s/Yes 2775*)
Yes small ruptured pipes or from -
1,0CTA IV cracks in large pipe which -
- actuate emergency core cooling _
I-
- Al-11
Uprating Input NRC Initial NSS Analysis Computer Approved hennal Power Assumptions AnalYSC1 Codes Vtili7ed OclR*Dcd[{l' OUtnut Assumed Used inadvertant loading of
- IIOPARD, Yes/Yes 2775 Yes a fuel assembly into an TURTIE unproper positmn u
yes Complete loss of forced reactor
- LOITRAN, Yes/Yes 2790 umlant flow FACTRAN,TlIINC Waste gas decay tank rupture ORIGEN2 No/No (h)
Yes Single RCCA withdrawal at ANC, Ti!INC Yes/Yes 2775*)
Yes full power Major nipture of pipes WCol3RAffRAC, Yes/Yes 2775 )
Yes 6
containing reactor coolant up
Major secondary system
- LOFTRAN, Yes/Yes 0(suberitical)"'
Yes pipe rupture up to and includmg Tl!INC double-ended rupture of a stemn pipe Major rupture of a main LOFTRAN Yes/Yes 2790 "
ies
- feedwater pipe 2775*)
Yes Steam generator tube rupture RCP shall seizure
- LOFTRAN, Yes/Yes 2846 e Yes d
(locLed rotor)
- FACTRAN, TillNC Fuellumdling ascident ORIGEN2 No/No 2831**
Yes Spectrum of RCCA ejection
- TWINKLli, Yes/Yes o and 2775*
- Yes accidents FACTRAN TlIINC Radiological TACT 5 No/No 2831*'
- GOTillC No/No 2775*)
Yes M
MSLil CThff IYf GOTI11C No/No 0,30,70 A 102%
Yes of2775t"
- a. Nominal pump heat of 10 MWt is assumed
- b. No pump heat (cor: thennal power) assumed.
- c.. Maximmn pump heat of 15 MWt is assumed
- d. Uprated NSSS power (including maximum pump heat) is increased by 2 percent.
- c. STDp with a TDF of 86,000 gal / min / leg assmned
- f. RTDp with a MMF of 87,800 gal / min / loop assumed
- g. More limiting tium + 7.0 pcm/*F.
+
- h. Tecimical Specincations limit on tank contents is assumed
- i. FSAR Chapter 6 contaimnent analysis.
W/rgm & SCS/ jaw It/14/97 Al - 12
NRC_ Question No. 9 Document what technical areas should be re-reviewed because current analysis is not bounding and identify to what extent the power uprate depends on current analysis. For cach analysis relied upon, identify the maximuu power level for which it is valid; For example, as discussed in question 1, you referenced the " initial licensing" SRP 5.2.2 analysis when discussing overpressure r.otection; however, you did not indicate if this analysis was done at the uprated conditions. Please perform a review to determine other technical areas that should be re-reviewed because the current analysis is not bounding.
SNC.Responsg No. 9 Subsequent to the Farley Uprate licensing submittal (SNC letter to NRC dated February 14, 1997),
the non-LOCA, LOCA and radiological analyses, the LOCA and MSLB containment pressure / temperature response analyses, and the NSSS/ BOP system and component evaluations / analyses were re-reviewed. In all cases the current analyses, or documented evaluations thereof, or the revised analysis performed for the uprate, support the proposed uprated power level and associated changes. Iloweves, the Farley designers did determine that the Component Cooling Water (CCW) flewrate and heat exchanger UA values used in the Residual lleat Removal (RilR) System fluid system analysis were incorrect. The cause of this error was informal commun ations between the FNP design organizations; the affected cooldown analysis calculat;ons have been revised. Re analysis with the correct input values slightly impacted the system response times associated with RCS cooldown using RilR System for the " normal" and
" single train" cooldown cases as presented in Section 4.1.4.3 of the NSSS Liccusing Report.
Therefore, page 4-12 of WCAP-14723 has been revised and is included as Attachment til to this letter.
W/rgm - l 1/14/97 A SNC/mge - 11/17/97 NRC Ourslion No,10 List all uprate assumptions, analyses, and analytical codes. Confirm that these are identified and incorporated in license conditions, technical specifications, or the FSAR, as appropriate. This includes ensuring that the FSAR documents all analyticci codes used in the uprate. For example, as discussed in question 4, the TS limits on Tm are not bounded by the values chosen for the accident analysis. The other uprate assumptions, analyses, and analytical codes should be reviewed by you to verify that they are appropriately incorporated into licensing basis documents.
SNC Response No.10 The methodolagy in WCAP-10263 established the ground rules and criteria for power uprate projects, including the broad categories that must be addressed, such as NSSS performance parameters, design transients, systems components, accidents and nuclear fuel, as well as the interfaces between the NSSS and the Balance nf Plant (BOP) fluid systems. Inherent in this methodology are key points that promote corre:tness, consistency, and licensability. The key points include the use of well-defined analysis input assumptions / parameter values, use of Al - 13
~
.g atTi a
-u
{
currently approved analytical techniques (e.g., methodologies and computer codes), and use of currently applicable licensing criteria and standards.
A comprehensive Analysis input Assumption (AIA) List was developed at the beginning of the -
~
Farley Power Uprate Project and was used to identify, consolidate and control the input
- - assumptions for power uprate analyses. The list included the specific parameters needed to license a core rated thennal power of 2775 MWt (NSSS power of 2785 MWt) and the parameters chosen -
i
. to enhance operational ficxibility at Farley. The AIA List was also used in conjunction with other documents (e.g., tables providing ECCS and AFW flow rates for various analyses cases), which provided additional detailed design and/or system performance data.
- Ihc following AIA Summary (Revision 0)1.as been prepared based on the Farley Power Uprate ~
Project AIA List (Revision 4) and is provided to summarize the key Farley uprate inayiscs input assumptions; Appiopriate assumptions will be incorporated into the Farley Technical i Specifications and/or FSAR. Applicable assumptions are already included in the supporting -
analyses calculations. The Chapter 15 analyses and principal computer codes are included in the 4 csponse to Question No. 8 above.
W/rgm A rs. I1/14/97 AI.14 -
i I
I POWEN L'PRATI'. l'ROJFLT. AN 61NM8 INPUT ANMl'MPTION M'MMAkY Rev 0 Date: 11/1197 d
Analyela input Aassumpe6am Pener Unrute Where Refereered haucce Analysts Value I uremuet '
FMR 1 est l'Isre 4
J.
NOMINAL 41INiiliIONR POWO.AI A 166 l} NSSS Poww. MWt.........
278$
T 15.1.'
f Owe Paww, Mwt flSb.
277)
T 1$ l.:
$ j.2$
PCWO.AIA.t 66
_,.m...
frWO.Al A.166 Net alent incid, MWt._
10 T l$.l.'
CN.TEL961, k0 t
1 Cahrimetric lnor Alkmaras(Cwe Power) %.... -
+ /. 2 S $$.1,2 CN.TA 9016710 t
RCP 'oww, Man,l.a4 MWI(rwwelAICA). m Sl;TSfool6$
RCP lest trua t. Mat. t sh. Inute O $1.
17 3e46 2-PCWO-A!A 166
- 2) RCA 11ermal floalen flow, Total, spm...........
2$t.000 T441 l'CWO.Al A-l66 R CN 11ermal liseum l' low. Pw I air. men..
46 No T$11 CN 1*L%1. ko &l l
CN.TEL96 26 RO 264.200 T441 T 3.21 kCS Meimum Measured ilow, Tclal.
RON Mutennun Measured flow Tatal, gpm(IS).-ppm (ruel DCA)....
263.400 264.200 T441 3)
T 2.21 CN.TSS426 ROAl HCN Mmamum Mesmated ilow, Per wpamp,. ppm r!$)lDCA)
- 27,1100 38.i00
- 88. t 00 CST *L%I. ko RCN Mimmum Menaurmillow. Per i Iow Measurmwed I hertarrdy, *6 0 'S)... spm (rmwe...
4 /. 2 I. + I. 2 4 7 4 4.l CN.1hb41 hp r
1 low Measurenwnt Orxertaniv.5(rmn-1OCA)..
e /. 2.4 T441 T 3.21 CN.TSL96 26 R0A1 F
ITWO.AI A 166 41 krN Mahanical lWan ilow. Per i ne. msn.
1019(m i
+
PCWO.AIA 166 l
$) RCS llat lat low nm 9$,300 0% SOTP Per law [e,r$mp.
I$%$O1favg).
PCWO.AIA.166 low s 20% Mi1 P (ocak). Per lac.gim z.
92.300 kCN tiest lat low l'CWO.AI A.166
(
mpm.
91JKo T441 RCs I6 cat t.nt
$67.2. $77 2 T441 T 2.21 PCWO.AIA 166
)
- 6) kCS Venne IFP Av Ternperature. F(f 8)..- : : x CN.TA 90167. R1 i
kCS Venne ll'P As 1emp. Man (w eme allow), F
$7).2. $23,2 CN.TA 90167 RI NCS Venne ilP As leme. Mm error allow), F..
$61.2. $712 RCS Teneerature f.rror Allemarav(w. F.
+ /. 6 51$ l.2 CN TSL%I. k0 PCWO.Al A. 66
- 7) RCS Venael IFP inlet Temperature, F-.
$30 6 541.1 7441 RCS Vessel ll P outlet Tengwature 15....
4 603.8 613 3 T441 PCWO.ALA. 66 2
.....4 RCN Vess.1 (LP Inlet Temperature, k.-,..m...
m
$47.0 PCWO.A1A. 66 CN.TA.90 167. R t RCS Vennel Upper llead Temperature. Man F. ~.,.-..
T. outlet CN.TA.90167. R I RCs Vml Uppet llead Tmuierneure. Mm. F T.6nlet PCWO.AI A 166
- )
RCS Ctwe lifP Asg Temperature. Fm
$71. f. $ k l.8 T441 PCWO. Al A 166 RCN Cave lli P inlet 'l ernperature. F,
930 6. $4I I T441 PCWO-AI A.166 y'
91 (4CN SH Outlet Temperature I'
$10 3 340 R
- 10) RCS Prenure Nominal, psig(TS)-
2233 T 4 4-1 T 2.21 PCWO.AI A.166 fu l'reasure, hmnal. gmig..---
2233 CN lCAT(96>7. R1 016 CN.TA 9016
)
Pn l'reasure, Man (o error allow twe).poig..
228$
CSTA 9+167 kI 2 INS T441
?n;'reneure, Mm (w erns allowarne), gaug..
4 /- $0 S1312 CN TSS.%1. ko vr Preuure f.rm Allowance, rui..
ll) RCS Yolunw Ambierd), Total H 0% avs SOTP ft',
9342.9 SI;i3 Foal A 0003 RCS Volunw Ambwed),1otal H I S% avg SOth, ft.o 9194 0 Si,TSECA1A4>003 RCS Volume Ambierd). Tutal R 20% :wak $(ITP, ft -
9077.7 Sl;TSECA1A4HX3 RCS Volunw Ilus lotaltm o% avg EdTP 11 9829.2 St.TSE-C.A1A4)o03 i
p RUS Volunw Ilot To'.t i IS% av SFsTSE-C-AI A4M)D3 11 v469 8 T $ l.1 p 'f.nAl 9330 0 RCs Volunw 'llot Train 2(
St.1 Sl;oAIA 0003 RCS Ateste Y'olune ( tei e me),
m, it.....,....
3291 513242.2 CN.TA.90 303. RI RC8 Active Volung (Star.up & I;ull Power), him,it'....-...
7627 S IS 2 4 2.4 CSTA 90-303, R t RCN Midl. e IIcatup Volume. It 949 See Note N70
- 12) RCS Mam at). Tot g o% avg SOTPA Tavg.$77.2F,ll...
Ala,809 SF.TSF C.A1A.0003 RUN Mass Tot 0 0% avg SOTPA Tavg-367.2F lbs.,,.
411.472 St./l St.oA1 A.0003 RCS Mans l ot Q l $% avg SO1 PA Tavg.$77.2I, ilu 394.637 S F.TSF.C.AI A4XKG l
RCS Maw
. Tot W I S% evc NOTPA Tave.$67 21,lts 396.049 SF.TSF.C. AI A NKH
- 13) Pu insel, Nominal 0% IWer. % span... o.
21.4 CS CAT 2
Pn iswl, Nominal 100% Power A 1 asg-$77.2F, % span..
$4 9 CN. CAT 2
l'n i4vol. Niemnal GOS Power A Tavy-$67.2F % span,..
43
- CS CA1 2
4 Pn twel Imt Allowarwe k8 0% Power, *e span.
+1. $
CN.TSS.El, R0 Ptr 1 evel I rror Allowarme d 100% Power *, nean.
+ 1. $
CS'lSS 41 RD I4) Steam flow,llwie per sos c.,. -
4 07e+6 4 09e e6 T 3.1 1 PCWO.A!A.166 Steam l'venaute (hG Outlet),
_a 67). 798 T $ l.1 PCWO-AIA 166 Steam Temperature, F..
.. --pmaa.
499I.$180 T$11 PCWO.AI A.166 M*1ure Carryo.vt. Mas. %.
0 23
$ $.1 I.
PCWO. il A.166 y
- 13) Main reed +atet Temperature, F... _.
4434 rt11 PCWO.A!A 166 Mam teedwater flow. lt=1w per 80, 4 07e+6 4 09e+6 T$11 POWO-AI A 166 Al - 15 5
---,-+n e, m, -, - m--& --m m va
.-n v ---.
.e w.-.m r e av-e a v +,,,,.
-m
..--me..--,e+
. e e --e e v ww.
w--
. ~ - - - - -
vv
Analyuls lerni A r l -
Peeee l' peace Meet liefensued heutte a
Asse6pais i aime linerumiest FMR Ted Emee CN CSA.94407
- 16) 6014 vel, 0% Powst, % NIL..~..~..-....
$4 $
CN.TA-95189 60 haeulary Neunianal Maan Tadald 0% Pima, He'SO.
Ntte 62 60 Isol,100% Poww. % hL.........~.-
CN CEA.94407 bO Em Nwnmal Maas. Tidal 4100*6 Pown A CN TA 93189 1 ave.577 2 He'hu.-..~.m.....~.........-.....~
Nide 62 f41 Ewa Namnal Mama, heal 4100% hmer A CN.TA 93.lk9 l a s p 367 2. Ilm $0.........-......m.....--
~o Note 62 CN1hs961,RD Su swl inw A'kmaswe $ 0% hmir, a6 NR.....
4 /. 7 CN ' ISM-96 l. ko
$O evel I me All<mance 9t 100% Pown. % NR.
4 4 7 w $ bias T/Il-9$ lH
- 17) Owe Avwagelleall1us thuie N';
197.200 T441 189.k20 T441 T!Ild$*133 Fil95l$1 Jt) Cie, Awenae lancar hm+rdw'fl
$ 49 T441 PCWO-At.A 166 19J Tenio I'me Allm6am e far $O I nulme i; O
PCWO-Al A 166 20 C.ve Ibsens flow (markTDPL a6 7i N4413 CNJARIJ:Y.00$
- 21) Cwe Prmure 1) rop, pi _...,. _
.m 204 4 20 7441 CNJARI.EY 00$
is 3 4/. I R T441 14.
14FAClOli(1801 ANT NWI Ehl l) l'kl3SURl7.f R
$1;1SF.C Al A 0003 Pu S)vay Valve Capauty (l 300 T $ $.14 Pn 1 otal Volunw it{celgn), yign), l'.acit gpn%alve...
m.
21,000 4
Pn l edal Volume (f il,
l400 t $.$.9
$1/1 hlW.AI A 0003 6F.ThlW.AIA4KKO 1422
-..... - _. _.... ~..... -.... _
Pn Rehef Valve depanty (Design). l.ad. llak/ valve e EI A 61..C.AI A4XKO
$lA%lW.AIA4KH3 Pn Enidy Valve Caputty(lleugn),1.ad, Hm1v4alve ~.
34 000 4
483 J S.$. 4
- 5344, hl;/1%i4A1 A 0003 Pn hafety Valve $ctpura, pig,.,
o.-...-..~
$34 CN.TA-90-167, N i
'n heMy Valve henpurd Tolerana, %(1N).
4!. l CN 1 A.96422, Ri
'n halety Valve Setemd hhift, %......
- 0 9/. l NML93-0113 on katet> Valve (mri Esal Purge linw, nes... ~......_...
1.6 CN ICAT(96).016 Va lleetw Capautv as inn 1 Tidal. Nw 1400 e
- 2) N11:AM 01.NERATORN
'CWO.Al Av 66 ku l ute Pluggmg, Mus % avgpcak...--..,... -..~.....
09
'CWO-AIAi 66 SO l uhe (1ugging, Ma A % avg peak..
~..~.m.~
I S'20 T441
... ~..
CWO-A1Av 66
$01ute Pluggmg, Ma% % avg peak (mm lDCA)m 20'20
'CWO.AlA. 66 NO 1 uhe Plumnt Mat % avg peak (I DCA).
20'20
- 3) RCN EUROl:AND $PR AY 1.lNI.8 El;TS' W.AI A 0003 HCN burge.ine heas Range (In Pn f.esheurek pig.,
22.15 M;TE.. C.A1 A4KKO ane Tong Range In Pn i.enheure),1...
600 633 RON Eurge ane hess kange ((in Pa bsi,murek ig -,..
2233 2100 El;ThR C.AI A4KMB NCENpray
$1.1 % iC. Al A4HO 4 kCN Nwav
.tne Ttmp Rante (in IV I swimureL M60 944 P,
M Al% NIFAhl ANllllTim All14 M Nil 414 l) MAIN sIFAM sVE11 M
$1,1 NIX.AI A.0027 1083 T 10 31 h0 shcu Demgn Ntearn heure. ig -...
12,9k4/4 0 Il 3 41.1
$1/1 SIT AIA4027 MNNV Capauty,lotal, Mm, thi TN).-.m.-
11 kk0,260 M%NV Capaniy, Total, Me,lh1w sum-IDCA)..
4 CN TA 90167. RI 113!47.1 si;TSE C.AI A 0027 MWV rapauty, l as h, Mm. lhk'veh e (TS)-.....~...-
kri$.644 M%V Capanty, l.aA, Mug th1w% ah e (rmelDCA)-
N38.684 CN.T A.94102, k i MS4V Scipimt, Valve I, pig
$)-
1075 T 3.74 EIA SIW.AIA4M)27 MuV 5etpunt Vahe 2, pig E ) m...-_
108k T373 M21 $lW.Al A.0027 T374 sl A SIW.AI A 0027 Mss V nettemt, Valw 3, pig 1 02 M%NV setpunt, Yalve 4, pig I
IS
'I373 EIA SI C.Al A 0027 T 3.7 3
$FJhlW AtA4027 MNSV setemd Yalve $[vk,
. ~....... ~........,
i I.29 MuV Accumulatuws Y 3M
~
3 CN.TA.94102, RI MS$V Anumulahon, Yalve 4, Mas b.,as, %.,.
CN TA 94102, Rt 2
CN.TA.94102, RI MSNV Anunmlathus Valve $, Man Imi 10 MssV luomkms Mas %...-...
10 CN-lOAT(96)015 m,...
MsNV setemd Tolerann, %
43 T373 St.Thl4.AI A 0027 ARV Capaaty at 1023 pig, h(TS).......--.
M;l%IT.AI A 0027 eal, thk -
122e+6 T 10.31 ARY httpuni, pia... -..
102$ to 108$
T 10 Fl
$14%I4AI A 0027
......... ~........
ARV Capauty at 102 $
403,$00 T 10J.!
AR V Capady at 10R $ pig, l.ad, faktvalve... _..,._
SFJSlW.AI A 0027
$10TSl: C.AI A 0027 pig. l uh. lhkeva lve _.,... -...,
619,000 EllCapadty at l00% Powct ( Tavg * $77.20 Tidal,lhk..
3.$0e 6 SM 934021 Ol4 SD Capaaty at 100% Powes & fasg a $67 2F. Total, this...
4 6Ae+6 SM 95 072'4)l4 hD Capanty at SOS Poww A Te is - 3712F, kat,lhk 6 33e +6
$M 95 0723 014 bD Cagwity at $0% Powee A bg = $67 2F, Total. Ihlw..-
S93e46 SM.93 0721014 ED Ecio.mta asg.Tre0 8't 41%mer ATng = 3752F, F.. a 9.3,15 I CN.lCAT 96) 023
- ~
$D Aetomda as.Trof M% Power Alng = $612F, f-7.0,10.1 CN-lCAT 23
$D $ctonnia ' avg. Tnt) 100% hmer A Tasc * $77 2F,1%
15 l,30 2 CN.ICAT 23 SD helpunta avg.1nt) 100*a Power A lase = $67.2F, F.
10.l.20.2 CNlCAT 23 Delta 1 50 Outlet to MSSVs. Mas pi(m+1EK'A).
20 CN.TA-022,ROAl Delta i SO OutM to MSSVs. pi..-~~.--....._.
10 5M.94 0721002 Unimlable Steamtme Volume. Per I cop.qnt, Per imp, pi..... -..
2$
Incita P Sd Outlet to $leam lies ler Mid SM 95 072 002 Il Kee Att 2 SM.94072.004
- 2) MAIN ITEDWATI R SYSTEM t
Ummlable reedinw Volume. Per Imts it'.
See Att 2 SM-95 07214WO Al.16 l
.m.-
Asialynne loput Annuaispinum Power Oprete
%)iere Hefertsi4ed hource Assalym6s % alue Datuesecat IMH Tesb Esiec lt RIVlOR PROll PllON NYN10M T221 CN.T4%I, R3 l} hmer Range. lua,liigh, % (11) m....-.-....
109 CN TA.5167 R0 hmw Range lut lbgh,'6 evviIDCA)...
l' 8 T l $. l.)
T 2.21 CN TSL%l. k3 Power kange lun.14m,%
S)..............
h3 CN4 A.90 277, R2
'ower Range llus-larw, % ruet.lOCA).......
M Tl$14 kley l enw. wtune 0h T l$ 14 CN TA 90-167. R0 T221 CN1SL 41,R3 2) l'uwer Range thrh Negative Rate % R11'nS)..
$0 Poww kange lbgh hartwe Rate,h kl P[lS).
$0 T 121 CN.Tbb%I kJ I kisy lime, acumuls..
P$
S7212 90AP' $400E T221 CN TSL%l R3
- 3) Intermediate Range Neutrem l'lut. % RT P (TN)..
25 0 90 AP'.LM*i
..+,..
Iklay 1 rme. mmis OL T221 CN 1SL%l R3 4)
Suurt, Range Neutr<a Ilus, ele fly......
le+$
90 AP*-54'n00 kauru ky,rse,lrip iklay 1 eme. wwe US
$) Ovw Tarviorature ikka-1 **etra. ire (1)...
Variable T
$ '.)
T 2.21 CN.*lSL4l, R2 Variable T: $
3 T221 CN TSL%I, R2 Over.bmer Ikha 1 betpomta D S1..
(011)Ql2 T
3 3
ikla y 1 rme, seetwala..
CN 1 A 96167 R1 CN TA 90167. R1 (OPDT) 12 T'
io T221 CN.TSS-%l R3
' rem.urver vessure. lam ik63
'ressuruer 'reasure l oc., pig (I F).. _......,
6)
CN.T A-90-167 RO 1823 T l$ l.3 1cnsure lbgh, pig (nic-IDCA)
T 2.21 CN 1 SL41. k1 pig O S 244$
1essuruer "rennuruer Wssure lbsts gang (rum 4)..........-
242)
T $l3 CN TA 90167, R0 klay 1 tme, neuguli..........,....,...... _..
20 t
$ l.3 CN.T A.46-022. R l klav l une tpreuure lhchi smme { nim.l oC A).
10 T $10 T221 CN.TSL41, k1 92 Tensuruer Water t sul.l hrh. % span O.S) ! DCA)..
90 AP* %4sf HiN 7)
CN.TA 90167. R0
'reasuruer Waier iewl 1kgi, % span (n c.
t IDo klay 1 tme, AmmM 20 T 2.2 1 CN.TSk%l, R3 ew llow 1rtp Sctpunt. %(I S)lDCA)................
90 k)
CN.TA.5167, R0 ew flow 'ine Scipunt, *o (twe k$
Tl$14 CN TA 5167. RO kla y 1 rme, wmA 10 71414 9) l ew? ou SO Water sul, % NR (TS)............. -..
2$
T221 CN TSS 41, R3 CN TA.4105 O
l$ l 3 sul. S NR w envuim allow),l),.
- ew-ew Mi Water CN.TA %04l esel % NR w>o envirim alkm.
16 15 l 3 l uwi ew M3 Water CNCRA 96 3(-
l mw ou Mi Water evtl.S NR w/o environ alkm).
13 CN.TA 90-167, R0 1 kla) l eme, omguin 20 T I$ l.3 T221 CN TSL%l.10
- 10) timlerwhage RCP, vohs on.h hus (TS)..,
26ko CN.TA.90 243 WO titulerwhage-RCP I klay 1 mw, teomits..
1.$
T221 CN TSLEl. k3 lirulerfrequerwy REP, is each bus (1 S).................
$ 7.0 CN.1 A 90 24S, R0 LInterirequency HCP, is ea$h lea (nim-lOCA)..,
$6 8 timterirequency RCP. klay Time,sm etii 06 CN T A.90 243. RO
- 11) R n 1 rip iklav on Turb Tnn. nco.
10 T IS.ld CN TA 90-243 la Auto mp Preu, ruigh S).
49 T221 T221 CN TSL96.l. R3
- 12) : R Neutrim I luu (P 6), amp Isl..._._......_.-.....
le 10 T 2.21 CN 1SL%I R3 ew hmer kmlin Ithst
' 7, P 10 Iw), % R1 P O S)..
10 ew hmer knlnpsliksk 7, P.13 L % Turb imp hess T221 CN.TSS 41, R3 ugu i s O S ).........-....,... _
10 Amrt Range Neutron Ilus (Pd_)......._
- -...% R1 P fl S).... _....
30 T221 CN.TSL%I, k3 T 2.21 CN-TSL961, R3 hmer Range Neutism ilus (P 10 Ikt). % R1 P O S). ~....
t T221 CN. l~S S-E l, R3 Tmh imp L liamber (P 13). % l urb imp Preu ! qmv (TS).-.
10 T221 CN.TSS-% I, R3 R s l nr= liksk, I uthm mg l urimie 1 rip (P 9), % R IT (1 S)_
30 Ikla) 1 imejeunt.
0$
Sec 7 212 L.
OtJy l(01.EYNTITIS l}
PklECR171 R CN lCA1(96 016 htywimnal Spray, On, pig..
2260 T $ 219 Propwinmal hprav. Maunumt pig -
2110 T$219 CN-ICA1(9( )l6 lladup ilentors, Gn. paig..
2210 T 5.219 CN ICA1(96 16 Rehet Vehe Setemit. pig _... _.....
2333 T S 219 CN-lC aTl 16 Rei ef Valw Ormung Mroke lime, Mas m..
10 CNICAT 023 Rchet Valse Climmg Stroke lime, Mas m..
6 CN-lCAT 23 Per Iswl,0% ILwer, % span..._._.,
CNICAT 2
Pts 14wl,100% hmer & Targ*$77 21'. % span,._.
21 4 CN-ICAT 2
$4 9 b71.twl.100% hmer & Tesc467 21' %inan 41 R CNICAT 12
- 2) SilsAM Ol;NFRATORS NO 1al,0% hmer, h N R,...,., _...... _
$4 $
CN-ICAT(94)-007
}O l ort, Orenter than2% Power. % NR.
38 CN lCAT(94 @07 Al - 17
. - ~ -. - -..
L 1
I Analpolslopui Aneumipteum l'emer l'oeiste
%here Nefereewed huerte i
Analysis % ator Inucumwal MAR Tedimmet t
- 3) ItOD CON 7k01.8YETLM CNJ CAT 96 416 i
Raul Cween Avg len p lingh I.hnn, F
.m...m
$67.2. $712 CN. CA196 423 Rod Cimaru Atg lerup lam lamst,l'.. -
$47 CN.;CA1 4 21 l
lirwar Unm- -
z-.......
ag Tenw Oswetenldreyamk sec..--
I med line Cawstard Meest m 40
..m..
CN. CA196 4 5
as 1inw Causenta Meml m...m..m 100,'00
,...m.mm....
'la
'owie Miamakh 14 !)ependent Usin --
.mc ft,8 CN. CAT 48 l'oww Mesmate Varsalde Oam =
CN.lCA1 4 16 Rod fysed, Mut, starvemn.
m.o.-..~
..m.
e/. f.)
keul heeed and Dumason Deadt asut I.
CN.hCA196 416 i
8.0 CN4CA196 416 kod I. peed, Mas stegvnun..
4{.0
$y2l1) 70 CN TA.
2: 7 kod hpaed, Mas, busi'veim.. _,._.-...-.....-..
m.
$ p.2.s.2.I CN.lCAT(W4016 kml(.marolImv for Mautemm Asieed Mas F 63 l',
F.NGINF.F.RF.it EAWTY FTAll'NFA l
- 1) S AIT.TY INJLC110N BYSTI M T334 CN.Tss 961, k1
$ Ad. C+ntauviwnt Pressure, llesh, pig is)....-
40 CN-l hs.961, R3
& Ad. Ceedanswed Iremeurt llish, paig i OCA)-m..
7O T 33 4 CN.1 hS.961, k3 1890 Ad. Procuruer heesure, l Am, ptg ] )}u.~%...
1700
...m..
CN TA.90-167 R0 Ad. Mcamlow Pressure.)14m, peg jl emet. A CA) e 537 Ad. laressuruer Pressure h
T324 CN.l$$+961 b b) AICA),.
CN4'k A.96.46, R t h
Am, pig i Lrum.f.-...........
Ad. Mesmluw l'ressure.14m.
h Ad. Meamluw Pressure, low, png CN.TA.90167 R0 fruwe OCA),. - m.
429 peig T 3 3-4 CN.1ss.941, k1 Ad. Ihnwadial Pressure lietweca htsamluws,(smd (16)..,
100 h
27 Ad klay 1 w all ite pmerL Mas, m trum.1O 'A) _.,..
h Ece Nde N22 Ad klayfw oltua imwn). Mag sec (lICA).-._..
I2 T7316 h Ad klay Wo olliste peer). Mas ses (rum.! A CA)....
See Note N22.
h Ad Iklay Wo adfane 27 T 7.316
$' Aduatum fkley, Mm.smak man en (i ACA)-
CN4'RA.96 23 ku esovuls (M fl R )..,.-.....m....
0 T 3.3 4 CN.Tk% 96. I ik3 I h r Inteelmk pressurwer hensure (P.ll), psig (18)..
2000 bl/l M'oA: A 4003 M
- ump C<mfiguratue.............~..
CllTI & RllR
......-...._~..m...
M;TEl4A: A.(KKH R Purnp l cad ikyadatum Allowarwe, h ikstgn llend...'M Pump lJIced lkyndatum Allowerse % lknign licad a...
10 Cll hisThE&A A4M103 10 Rll hl/l M4A1A-0003 Cl' %I truedkm llow inenlarne Alkmasus, spn........
20 see Attachnwed 1 M
low Rate (Mm Kafeguards), Mm. ppm vs psia -.-...
see Att i F 1$.2 41 See Attaciunent f El low Rate (Mas Enfeguardat Mas grun vs sein
.mm
$ce Att l EM.94 0453410 Time of No to C Recus Mmilafegua: Ink him, m...
.. 4 2139
$M.94 04324M)
Tmw of Solu O.Hwtre Mas Nefe atda t Mm, sw m.-m..
13ll See Note N69.
16nw of AOlaC Re6uc Man Enf dvno kilR k Mm sec.
2400 See Note N72 INrathm of Rilk Intwrvidum (Mm afegda). Ma% see..
IB0 See Note N72 theatum of kilR lederruptum (Mas safegdi), Mag men.
l*0 / 330 El C.A All.210101 15 1 mw of EU to lit. Rwuc, hro_..
..m..
CN.CR A.9&l 4. R 0 El 1 ernp ha mg Recus, Mag l'..
..~.~......._.~ 4 207 El.C.N All.2101.Cl M 1emp hutog Recuc. Man.Asg (0 2 $ 000 awk f.
190 sM.9%0721421 M Temn Wrme Res uc. F Ree Att 2
- 2) R t.lUI-ljN(I WATI R STORAOL TANE ELTSI:CAlA.0003 rwa l 1 dal Volume, mal -. -
500,000 T636
..........~...............m.....-....,
S31.26
$171 El4AI A4KiO3 rat 15 471.000 RWST llehverable Vol(Mmica 1. 2,3, A 4), Min'I'k),(....-)..
E 31.2 3 SLthleAIA4001 l' WET Ikhverable Vol(Mmics S A 6), Mut gal (
30.000 S31.26 si;TM4AlA400.1 2100 kWET lhme Cinweidemism, Mm. ppm ((TN I...
2500 S31.26 ElWEL&AlA4M103 RWNT ikom Curweensatum. Mas poru TN)..
$00-NAll.$024 00 RWNT 1miperature, Nunut. I F (IDCA) 33 RWNT 1 enterature, M ut id 5).......,........
N 3.1.2 6 Tech Spw
._m 3$
CN.TA 90167 R0 RWE T lauperature, Ma% I (rum-l OCA)...._.
120 SI C.S All 302/C0 kW"T lenterature, Realutic Ma% ) (llCA)...._._...
100 hM 944432 001 RWST Volunw at Ibph leel Alarut Mag gal....-.
$06,MM)
SM 9444924M);
RWN l' ikhwepble \\ ol M ho la Cl. Rentd. Enl 121.000 31 llokle ACID $10RAOl; TANK N 31.2 6 SLTSECA1A4XMd ltANT Volunw Mmica l. 2,3 & 4), Mm. gl(TN)-
m.,..
7(MM) 11316 N 312.$
M/l SECA1A4MMB
'lANT Volunw Males S & 6L Mm. gal 2000 E 31.2 6 EL1 NFCAl A4MMO iAE I tkne Concmaratum, Matt plan N 31.2 6 hbl NE-C.AI A4M)03 lAN f Ikom Coswearstum, Mas ppm IWT l emperature. Mut ITIN) 7700 N 31.2 6 M.TM -0. Al.A of Ha 65 Al -.18
~..
~..
Asimlyula leapest 4.aa 1*La Pomer l' Mineet iteferseceal naenete Asial ste 'alue Dareusseet i
FMAR Terb Assee 1
- 4) ACCUMULA10RS
$r.TsE&A1A4003 Aw Cteifismnuars Np/Acc.~.m.hA 1450 T 6.31 3
ml er.<-
SF/150.C. AIA.0003 i
Acc ltdalhdunw. A
$1/l SECAIA.0003 102$
Aw Water Wlwew.N manal ft w....
.o_.--
S3$'
M;TSFSA1A4003 he Watet %dene, Mn gal ris).m..-.
15$$
8 3.SR
$l;1El;eA1A 000)
Au Water Volunw. Mas sal (15)......... ~ ;
7780 Aw : keun Casugeernuin Mn prun 15) -
2200 T 6.31 S 3.5J St.TSIAAlA4003 See Note N42 Ac : keun C4sgenernines Mm lgen (DCA).=--
2l00 CN.TA.90167, R0 Aw turse Caswenernenot Mm, ppm *1DCA) x 1900 Aw Ikwie 04 uwentstum, Mas ppm ' ' )....-
2Mio T 6.31 8 3.$ 1 M;TSl4AlA4003 El'C.SAll.$024C0 Acc Water Tmgeretwo, Nnmmal. F AK'A)~m 10$
sUlsl4A1A 0003 Aw Watw longe,eatare, Mn F...
60 T 6J!.1
-...... ~. ~..
CN.TA 90167 R0 Aw Water Teegwature, Mat F.(two-1 A CA)..-.. :
120 51.C.$ All402400 Aoc Water losgerature. Reahshs Ma% F (lDCA)~..o 120
$LTNr C.A1A4003 M
T631 Aus Oas 1 emper etwo, Msg F -
Acc una Tergerneure, Man, F (rwe> l A cab. -
l20 M.C.SA!!40J/.C0 Acc uaalongwenewe koaluuc Mas f (IICA)-
120 Aoc hensure, Normnalmig (1 A CA) 625
$lf RAll.$024 00 8 3 $.1
$1/1Er4AI A 0003 Aw Preeaure, Mm, pigTI5; 601 M/l blW.AI A 0003 w errur al owarus) Mui, pig (11CA).....
38$
T 15 4 2 Acc *iemaure (Mat. pai (1 E)...-.
S 3.5.1 MLTNibC.AIA4003
r easur e, 649 Aw:
Ac 4caeure f w eerut flawarret Mat. rnicil DEAL M,$
M;TEf C.AI A.0003 l
$) CllAku!NO PlfMFN
$r.151:.C.Al A 0003 Chargmg INmp flow. Desip, spin..
..m...
!$0 T 6.31 kl.TsroAlM001 Charruw hmus ihm. htat men,
650 T631 6)
RI.ElDttAl. Ill A" RI.MOVAl.SYSTI:M
$F,TSf4A1A4003 37$0 T $.S.8 kilR IVmp liow. 3ceip, men....m.io_w...
.m...n, it !" uhr...._.
60 Be+6 T $.$.7
.. ~. _..... -...
M/fsr4A1A 0003 Dway lleal v4 20 hiurs After Shuh EF.TSIZ.Al A.0003 kilR C4=,hlown Cut 4n pressure. psig...
- 425 T$S7 RilH Cuohl.mn Cut 4n l es@crature, F....
..~..
- 330 t $.S 7 M,TM4A1A.0003 SI/FNFOAl A.0101 RilR Coulewn Cut tn Tenw inus (NOTR).
2$
CN.TA 90167, it0 60 N l$.2 8 2 )
Al'W Attualmn Delay. namnds..-.....
Cale M 38 04 A 40 02 Al W low Mm. grun (emw>l A C A...............
See Att. 2 5 IS 2.8.2 I Calc M 38 04 A 40.02 Al W low, M m, span (IDCA sma 1 leeak)..
Nee Att 2 Cale M 38 04 A 40 02 Al'W low, Mag gren Design 'i rarwwma)...
hee Att 2 Cats M 38 04 4 40 02 See Att. 2 AfWL hm. Mas gg.,a rawlDCA).m.m....m.............
Calc M 38 04 A 40 02 Al W ihm va hena, h m, ppm ta. pau (ruel4 K'A),..
See Att. 2 Calc M 38 04 A 40 02 Alv I km n Presa Mat gun vi rma (runt A K'A)...
See Alt. 2 AFW 1 emper ature, khn, I (niel dCA)...
3$
Calc M 38 04 A 40 02 e..
SM.95 072140$
Ito s l$ 4 2.2.2 AFW 1 weature, Mas F....m-...
26ko T 3.3 4 CN.TNS 961, R3 Al W kcl I halen uhage, vehn (T5).....~~.................~.._.._
AFW l rge Volunw hior to Dchvery of Cold ATW, Man Per Nmrmramm ne'a b7is n' W i
- )
ST0AMijNE ISOLA 110N MSIV Stroke Time, Total, Mas nec 1 N)....
1 E 4.7.l.$
See Note N48 CN-TA 90167 R0 MSIV Stroke Tmw, Total. Mas sec elDCA...
10 MSlV Stroke Tmw, Total Mm, nec ( DCA MA):).. -...
See Note N48.
0 Etcamhne l sul l mw. Total. Mat we -............,....
9 T 7.316 See Note N48-steamluw lmd Tmw,1 otal. Mas, see (ruelDCA)._~ __
82 CN.TA.bl67 RO See Note N48.
. sol Tmu, Total)Mm. we (lDCA mal')..
Nteamhew 0
T3.34 CN.Tss-96 l R3 wiftnd press. hgh Ihgh, paig (TS)....._.....
16.2 htcamfuw CN.TSS.%l R3 h)(,..~....,ip n,wlDCA)...
19 2 Steamhne wl.Ctmt Presa,lhgh-lhgh, T3.34 CN.'I ES-961, R3
$8$
Steamluw md-Nicam Prena, am, paig CN.CRA.%.I6, R1 kteamineLmd-Stenen 'resa, am psig ruelf K'Ak...
$37 heenmhne 'md.Nicam tesa, nw; t=ir tewel A K'Ai -
429 CN TA 00167. No 9)-
TURitlNE Trip AND FFEDWATI R ISOLATION lm CN lCAT(96)7, RI 023 l eedwater isol.RCN Avg l enniersture, Total,F....~,,..
$$4 lentwater Control Valw htro[e 1 mw, Mu acc~...
CN.TA.%I6 i eclwater Imd Tinw (kw ICV, Total. Mas ace, 7
CN.TA.bl67 R t i entwater ind Valw NtnAe 1)mw Total. Mas sec...
30 CN.CRA.%l2 Teatwater Imd Tinw (hw IWlvi, h4al, Mas see,.
CN.CRA.%I2 32 T334 CN.TSS.%1. R3
$O level. Ih h : hgh, n. NR (18)lDCA).._... _...,
78.3 SO14wlth : hph, % NR (me 100 CN.TA 90-243 lurbuw inp te av i olkmmg ihgh.lbeh 50 level, we...
2.$
CN.TA 90-243 1mbnw 1 no inela' foHowmr Reamw 1 rip, see 2S CN.CR A.0&t6 y
- 60) i LLSil Ul NI.kATORS l hemel uciwtativ Start 1 mw. nedfS)...-......... _..
12 0 54*l.1 Tesh 8 pee CN.T A.%i 67. R0 l
t meus oe mi, stan 1 nw. mnon-n 30A) n0 Al - 19
Analysis input Aasiasepth.m Power l'prote Wlict, Refeermed hource Analp6s i alue
- hicunient iMR 1et b Epte U.
( U R F.
- 1) IUllTYPr ANDIT.A1UR13 CNFD
'hpe._........
17 s l7 V)
T431 CNI't)
D'Ms (Y en %) m...
... ~...... -........
Yes T431 11unMe Plugs Renwved (Yem%)._..
Yoho CNI D I uel Rod Outside thanwtet. Irdies..
0 160 CNFD T 4 34.1 CNI'D T43
-. ~...
6 Ziratuum.
Number af Unda%laterial..
2 leviel and 3 IIM CNi'D Clad M staint............ _...._..,
Z1R107 re 4 CNID Auni litad ets (Yes%) _
Yes %
- 3) M)WI R Dis t RIUirl10N CONTROL.
$ 3 2.1 CNfD Asial Oflu4 Ctwerol Straterv RAOC AO Operatmg lland (r4 look,Cl 8)... -
~.~.I S)...
424430 832.1 CNfD rown 49112 AO Operstme lland 4 $0* Poww. k.% [6)..
E12i CM1i All) l 4msta fl S)......-........
' (1 I321 CNI'D Varymg CNI D Pv.ict Miape (run IDCA)..
9.933% AO 81312.1 l.70 (V N432.26 S323 CFID I Dil, Man (1 k)..
1.30 (lDPA 843.226 832.3 CNfD l'Dil Muhiptwt Cl N)....
O.
$4.3226 8323 CNI D N42.32 Tech $
Dil Utwtamfy (TN) 1 04 e
N 3 2.2 Tech 5 ievCO!.R Q, Mas (I bl..
2 3(Y N 3 2.2 Tech 8pwCOLR 2 32 (IDPA 1.081 S4223 Tech kpec p tlawtamiy(18 )...
S4222 Tech hpevCOI.R Is surveillaine Requwenwed (IS)...
I-I A U) Curu ~ E1......
11at k(
l'4321 l'3.22 Tech hpevCOLR CNI D Aane f uellies iL udwe. -..,
143 T431 N124 Tesh Spec WIR atmv 5t a h wcr (18).
t 02
- 3) Tl MPI R A1URI.1I l DilACK CON 1 ROI.
At i C 1 callmL (llOl. liZP), Mm. pcmT (1 S)-.
47 0 S4312 S3113 Tech Spec RSAC M DC (111 P), M m, d K'rm'oc............... _ -.............
00 TIS l 2A
$ 31.1.3 Teth SpevCol R MDC (I OL ill r ARl), Mas dk')gmtc..M I C fealla k 1101. Til P ARO, Mat pctn I (TS)..
Variable RSAC U.S 11516 CN.TA-9t. 0$0 kett re t emp at 1000 paia..
......~......m.
Variatile
.' l$ 2 40 I
l eth Spec 300 plwn til P Survedlam e (1 S).. -
Variable S 4. l.1.3 1ech SpevCOLR l hyplev Power I telliciend, pcm%P..
Variable F 1313 RSAC RSAC l hyplev les ap Cucil i eedtak, Mm, pcmT.
09I RSAC lh gt e leng Cocil i eedimL, Ma% pcmT..
30 l
See Nate Nio.
Des a y l lent..........~...........-...
Yarpn l l$ l.6 Muumum 6 w it lie Man l'url Tetup..
22 I CN1'D CNil) l usi Avoage lemp vs kwit -
Varymg CN)D i uct As g 1 emp tincertamty _.......
Yarpng CNill I uel Centerluw Temp vm k w'fl..
Yarung CNil) fuel bus f ace l emp vs kw 'ft-.....-...........
Vemng S31.14 NTD.NSA.TA 95 330 Muumum Temp for Crdicahty F (I S).............
541 t ow Power 1%urs 1 cat RC%ilm Temp. I (1 N) 311 N 110 3 4)
Till RMAl,llYDR AUI.lCS. DNil CNID llNil Ciwe 1.smd T Inlet n Poww..
Varyuig OTDT Neferetme Aual Power Shape -
1 $$ comme
$ 4 4 3.2.2 1121.1 CNI D l'211 CN.1 A.96 011. RI DNli Nahty 1 muts (I N1......
Varymg CNID Aua! Ottwt at lanut DNilk._
Varung DNil Cistelatmn..
% Rib 2 T 4 4-1 CN)D CNI'D WRik t T441 Mmmium DNitR, Niuninal Cimhtiism..
2 21(VS T441 CNI D 102 (1 OP AR)
)
T441 CN!'D
i np Reaamly Wivth,
- IMta L 48 S l$ 1.5 RSAC
'I nn Readivd v Shapo...............-
Variable F1513 RSAC RCCA Dron Tmic, neowuli(TS)....
27
$ 151.3 S3134 Tech Spec Maumum lhygel HCCA Worth, gun........
Varpng T I$ 412 RsAC R$AC 1000 Maumum i. ected RCCA Wirth,4 1)cha4......
Varying T IS 412 Maunumi l-eded ROCA I RSAC SDM(Ma 3), *61 kita.L -.....
t0 S31.1.2 Tech Spec p..........
hDM (Maku I,2. 3 4L
- Delta-K (TS)......
I 77 T 15 4 22 S31.1.1 RSAC Rance ol' kCCA Pad til Positisma, steps (I E)..
223 2.11 S31.34 CNID katInnatum Omuts(TS)..
Yarg mr F311 Tech Spec D flevenhal Rmt Wieth, Mas pcm'ush...
147 T 15.21 RSAC Rod Wahdrawal Rale, klat ptmkmal....
100 RSAC H mi inserthwi Ratj, hln tsmwnf i10 R$AC Al - 20
i 1
Amelysse loput Aaemageous Pos.ee l'ntate Mer, Re4etenced Sumere Amelyu6s talme inesimmeet FMAR Test itsee
- 6) M!soliAANI OUs ellARAC11 RISTICS R$AC ltssa t.flettve, Mmlsea pedive, Mn IW)Cle...._.
0 0072 kBAC t
80CRodIW M)..
~
0.00$4 815.462.2 kBAC 0004$
4
~ lista Lilutve. Mui IOC)denent, w%.e m.
w.
S0 T esti S CNil)iwe h4aamman Alkm Dru
..m.
fumher af Fuel Rods per Annenddy...........
-.... ~
264 akate Dilutum Fkweste, Refuehng, inhut&mu and $tartup.
CN.TA.90403, RI 300 S IS.2.4 2.3 saun......... m.
.....-..m...._..........
CN.TA.90 303 kl lheta lhhne llowtate. at Powst 300
$ 13 2.4.2.4 Cetikel thstmill/.P, Akl.1, NOMglen..... &, IM O..
1800 REAC RSAC 1730 Osnal tkv<m (C1.P. ARI. NOXI; 1400)mc Van CN.TA.90 303 Difisraniel thete Wcetl6 8an'ptun Peak Rod laurem IU.....~..~........_.
62.000
'RFAC.ALMid INw P
M WI... m.
3$
~~~
Aggvasunate Cydp M% bower.
m CNill Curi)
.iiurnur uwostiti._....~,.
20.000
. 4puldenun Cycle Iturvue. MWD'M111.-.........;..
..m...
20.000
$ 4.12 i CNil)
CN.TA.96 094
- a6ed koke % Itmis in l> Nil ( Analyus lJmit), E.,..
,m...
20 kl'.AC.Al A.014
, aded I uel I tamm ( Aamumed1 %.
1
.Il-RITl4r.1811er.RM Al, DF.NION I*ItO( *FlWMF.
ROIMt)NTROL 1) lavr Li avrisinty. {Favg) '
CN.TEL%1. R0 47 t
2)
I'RI.$$1rkly.l.R lieuunser Control 14ce mir ff*).
221%
T441 Y 2 St PCWO. Al A.166 3)~ PRI Cl%10N RC5 CAIDRIMF.TRIC IfWO.Al A166 i cedwster T empsestors. F.
..m..-.~
441 4 T $ l.1 l'oedwales langes ature litur, % pan...... _.m.......
,*D CN.ThS.%
A cedwnier 'reasure, smia.
._.m.-
Pstm*100 4
m.,
CN.18$.%.l66 R0 eedwatet 'ressure Inst. pi...
+/.60
'ecdm atar ' low. ltde %O..........~
4 07et6 4 09e*6 PC%%A!A CN.1NS-961 R0 4/ I 9 eedaeter 1 low Ime, % dp spaL IT%hAlAl66
...m.
... ~.
673 798 hienm h euvre.1ma...-.-..._. ~....
- Etam 14cesuie Fme, % span.....m
+1 2.4
?CWO.Af A.: 66 llot i sg Tengatalure. F...
603 I. 613 3
- 'CWO.A A 66 M0 6. $41.1 Cold t sg l empnature. I'...~ -.
Piemuruer Pressurei msa (TN) 22so T441 T221
- TwnA A. 66
- 4) ()AILY M)WI.R CAIDRIMETRIC M.'WO Al A166 Iced *ates Temperature, F =
441 4 T 3.1 1 CN.TSS %I. R0 f eedw ater Pressure, pia........mm......
P stm + 100 PCWO.A! A' 66 l eedenter llow(Verwun), ibalv/SO =
4 07e+6 4 09e 4 T $ l.I PCunAIN 66 kicam litenne. paia 673 198 T $ l.1 1.
(1WTAINMr.NT ANilCONTAINMF.NT NYN1FMM O CONTAINMF.NP Cale 1106
. C ndanmwnt Net f ree Volume. Max. It (IDCA).....
.m...
2,150.000 T LS 4 4 See Nide N7.t Imtini Prennwe. A us psia la CA).
14 7 T
$.4 4
.-......_...m..m..,
90 T
$.4 4 SM.95-0721005 Imtiallrwide Temperatwo llulk Avenge), Mm,l'..(~IDCALmm 20 T $44 SM 93 072100$
M F (I OCAl..
.-.m._
See Att. 4 T' 544 Outude l engunture'Hddneta and Area. Mn (IDCA).-
See Note N61.
StruduralIlut Emk Nerudusallleet kmk Matenal14otwrtwo(loCA)-
hee Att 4 T 344 SM.0LOl 21001
- 2) CONTAINMI.NT SPRAY SYST) M CAF.NF 1045 2
T 15.4-4 Ostatmg Pungw, Mn....m...
CAF.NF 104$
i 2773 T ll.4-4 CAF.NF 1045 S way 1%w. MaA arm per pu qqlDCA)..
..m...
$500 T1544 heray I hm, Total,Tlas, spi inca) --,..... -..... -...
Caio 13 07 I ull 11ow Iklay Tmw i w o s-ite powerL Mm, see 1 OCA)..
4W T I$.4J Full I kna lblay Time Jw attute powerk Elm, nec (1(4 C A)...
Calc M 17A 2 32 T 3.3 4 CN.TNS.%), R 3 Epray Actand Preas t ligtelhgh lhgh). pmg (TS)....-..._.
27 CN 185.%I. R1
$ prey Ado ni Press ( balelhalet hcht enie (non.1 OC A) 30
- 3) CONTAINMt.NT FAN COOT.ERS 5M.lss91500 002
= tml of f.IIllcal Rennwal((w ollute powerL Kim. mec (IDCA)............-.m..
4 TI$44 operat mg i aus. M n...._..........._.....,._,,
Imt of 1.n llent Rennwal w<o ollute powett Mm, see (IDCA) 274 T ll.4 4 Cale 27.04 I an iteat Renuwai Capauty, Mas It'IUhr Tan 0
Cale 27 04
- l an lleal Rennwal Capacity (pulkkmuk Mas ti{l OCA)...~..
SM 13 291500402 luhr'l an 93 4ke+6 T6M
- ( L UCA)....-_...,....,.- _..~.....,
- 4) CONTAINMF.NT PURUE ISOI A FlON l.
FSD A 181001 Contanswnt huge Yah, twlMam Time. Mas sw.
6$
Al-21
)
\\
Analyuls impus Assesmytease Pomee 0 where Refecemeed Euerte Ameissia ' sine lhwensestet THE Ted Rect J. - (18EMICAI, AND VOISME(nWikOi,NV8 TEM KrA SE-C.Al A4003 Charging 1%=. Mint aseo 23 See Nite NS7.
I Charsms ihm. Nmnmal. gen.. z t0$
.m....
Our i hm. M an. gen.........,...............-....
120 see Nede N$7.
Et, Nite N37 RCP - levathe ihm (uso kC$ k N<mimal. Tidal. ges..
l$
SIASIT AIA 0003
}
Chas sms ihm Torigwrature egen 11:inlett F._
130 bl.T$lW.AIAun03 therging liow 1engaaiure e its annett r 4:$
$FA%14AIA4KC3 latamti heasure C4earol Y
- PCV 14 Sl $ctpund. leig..
237 430 ELA30-C.AIA4KKO l
Mised lied Demmeraluer Remin (dume. ft 30 Came lied Demwwratuar Remm Volunw, A' 20 al.TSl4ALA4KKO
)
$14$f C.AIA4K103 Purtfknuim ikrw Rate. Nienmal. gen.
)20
$lA314A1 A 0003 133 Puninatuilhq Rate Man, men 300
&lA SIW.A!A4KKO VCT Vaduine'ortuna%.~... -.. --
12 6l sF.TsIW.Al A 0003 ft.-..
bl.1514A!A 0003 VUIIsvel,b 30
$FA'$f4AIA&KO t
VC4 l emperature, hcenmal. F.
1l VCT Pressure. N<wiunal, pag.--
5 TIA'50-C.AIA4003 VCT Vapur Purse kata, afm..-...........-
01
$f.TSE C.AIA 0003 th4=vnie Denun I.flettve Cut in ikvim Ownentration, plwn..
' na S41.2.2
$1JEFAAl.A 0003 14nstgarwy thenium i kmvate to RCS Mm. gen [I $1mm ~~
30 Roswie 164 hap lank Deiliguratuvi. b' umber._
3 D 873039 Ell.4 It167266 Fervt le Ib Alue Tar & i ntal Volume, aAhmed 2 R.000 J.
(1WIPONI.NT (DOI.ING WAT VN MYI FM CNN. 0047 CCW l i s i!A. M in......................_.._..~...
3 De+6 CN.96 0047 CCW llow to CCW lis. Normal Cam 44.wn gulls... m..
7000 T9.26 CN.96-0047 CCW llow to CCW iIs. hingle 1 rain Coollmn, gwnlis.
7220 T 9.74 EW lhm to CCW lit. Dempi. gwnlis......._.
10000 T9.23 CN.96 0047 CCW thm to RllR l lu. Nurmal C<mlamn. wwnliu......
4160 T9.24 CN-964047 CCW kuvta kilk lu. 58nr.le Train coniamn. gwnlis m...
4349 T924 CN.964047 CCW 1s Aus lleal maa. Normal Omhken litulv........
23 80e+6 7927 CCW 1: Ausllent nada. Emgle Tram Coulden, lituhr..
1R60e46 T 9,2 7 CN 96 0041 CN46 0047 -
S W 'ienverature. Mas, F-95/973 l
WAKil. PHO(T.nlNO NYN1 VMM if213155 Oas Detay Tank Volume. fibra -
600 M.
10klilNI; UENS R 410N PCWO.APR 160 htsam Pensure (SO ( Adles), ik.at laimate, pia..
787 PCWO APR&>0 Steam lempurature, llcat I.siimate, I..
$16.3 l'CWO.APR 160 l
Mouture Cart 3over, liest i.atunate. N...
OI r
Al - 22
.<-m
~,, - - - -..
--.,.~w~,._..-
_ _... -. -.. ~... ~. _. _, - - - - - - -, -
NRC Qt!cstien No. II Describe how the steam generator water level bias of-5.0% is modeled in the accident analysis and where the bias comes from. Describe the effect on the tesults of nxxleting the steam generator level in this manner. (WCAP 14723, Section 6 2)
SNC Rememe No.11
%c -5.0% of steam generator water lcrel span bias represents the velocity effect on the lower narrow range steam generator water level instrument tap. The location of the lower narrow range instmment tap is in a high flow region (i.e., transition cone), and the effect of the water velocity on the water level indication is to indicate lower than actual water level.
The -5% steam generator water level bias corresponds to the quantified difTerence between the actual steam generator level and the indicated level. This differs from typical uncertainties wherein a window or range for the actual level, compared to the indicated level, is determined. For the Farley plant uprate, the -5% bias was quantified via calculation for the Farley plant in the same nunner as was previously approved for the plant steam generator level tap rekication, The bias was directly accounted for in the safety analyses as follout For those transient analyses which are not sensitive to the initial steam generator level (e g., RCCA withdrawal at power, RCS depressurization), initial steam generator mass is modeled consistent with the steam generator program level (which is equivalent to indicated level). The level uncertainty and bias are not modeled since they have no significant effect on the analysis results.
For those transients or conditions where a maximum initial steam generator level is conservative (e g., steamline break M/E releases to inside containment), the analyses nuxici an initial steam generator mass corresponding to the program level plus the measurement / indication uncertainty plus the bias.
For those transients or conditions where a minimum initial steam generator level is conservative (e g., feedline break), the analyses generally model an initial sicam generator level corresponding to the program level minus the measurement / indication uncertainty, This is an overly conservative assumption, used as a convenience, since the analyses may credit the quantified difference between the actu,1 steam generator level and the indication level (-5% bias) Note for the feedline break event, the actual steam generator level is modeled in the analysis / evaluation.
W/gls & wm 11/12/97 MRC Qttestien No.12 he plant specific modeling and analysis for the large break loss-of-coolant accident (LOCA) is not providal. P! case provide additional infonnation regarding the analysis assumptions for the best estimate large break LOCA calculations. Information needed is the assumed initial conditions, the limiting transient progression with discussion of why it is the limiting transient, Al - 23
single failure assumptions, loss-of offsite power assumptions, time step assumptions, and major plant parameters with uncertainties. Show that the calculations were performed with the approved version of WCOBRA/ TRAC MOD 7A, revision I, and provide information that shows compliance with the code limitations and restrictions, (WCAP 14723, Section 6 3)
SRC.RcFermcEL12 See Attachment 11 of this letter.
NBf_QOcsLipn No 13
'Ihc submittal indicates that the operating ranges for major plant parameter assumptions will be documentut in Chapter 15 of the FSAR. Please provide a sununary of the parameters and the assumed operating ranges where the analyses is valid (WCAPl4723, Section 6-4)
SHC.hesp2ms No.11
'the requested Ill10CA parameter summary including operating ranges is provided by Table 13 1 and Figure 13 1 which follow.
SNC4nge.11/1R/97 Al - 24
Table 131 Plant Operating Range Allowed by the Best Estimate Large Break LOCA Analysis (Farley Units 1/2)
Parameter Oprating Range 1.0 Plant Physical Description a)
Dimensions No in-board asumbly grid deformation dering LOCA + SSE b)
Flow resistance N/A c)
Pressuriur locadon N/A d).
Hot assembly location Anywhere in core e)
Hot assembly type Fresh 17X17 V3, no restriction.
Zire 4 or ZIRLON cladding f)
SO tubr. plugging level 520%
g)
Fuel asumbly type Vantage 5 Zirc 4 or ZIRLOW cladding,1.5X IFBA, LOPAR 2.0 Plant Initial Operating Conditions 2.1 Reactor Power a)
Core avg linear heat rate Based on core power $ 102% of 2775 MWt b)
Peak linear heat rate Fo 5 2.5 c)
Hot rod average linear heat rate F, s 1.7 d)
Hot asumbly average linear P
s 1,7/1.04 u
heat rate
- e)
Hot asumbly peak linear heat Fom 5 2.5/1.04 ram
- Note that the P limit is a maximum value. For purposes of core design calculations or in core measurements, the maximum value must be reduced by an additional 4%, yielding a value of F s 1.7/l.08 =
u 1.574 Al - 25
Table 131 (Cont.)
Plant Operating Range Allowed by the Best Esthmate Large Break LOCA Analysis (Farley Units U2)
Parameter Operating Range l
O Axial power dist (PBOT. PMID)
Figure 131 g)
Low power region relat've power 0.2 s PLOW s 0.8 (PLOW) h)
Hot rod bumup 5 75000 MWD /MTU, lead rod
- i)
Prior operating history All normal operating histories j)
MTC s b. HFP k)
HFP boron Normal letdown (800 ppm) 2.2 Fluid Conditions a)
T,,,
567.2
- 6*F 5 T,,, s $77.2 2 6'F b)
Pressurizer pressu.e P a = 2250 psia * $0 psi c) loop flow 2 86,000 gpm/ loop d)
Tm Current upper internals e)
Pressurizer level Nominal level, automatic control O
Accumulator temperature 90 s T.,, s 120'F g)
Accumulator presswe 600 $ P,, s 680 psia 3
h)
Accumulator volume (tank only)
%$ $ V,,, s 995 ft i)
Accumulator fL/D Current line corfiguration j)
Minimum accumulator boron 2 2l00 ppm
' Based on Generic BE LBLOCA Studies Al - 26
- = -.
I P
Table 131 (Cont.)
Plant Operating Range AUowed by the Best Estiniste Large Break LOCA Analysis (Farley Unlis 1/2)
Paranneter Operating Rance 3.0 Accident Boundary Conditions a)
Break location N/A 1
b)
Break type N/A c)
Break site N/A d)
Offsite power On or Off c)
Safety injection flow 2 values utsd in reference case
()
Safety injection temperature 70'F* 5 St Temp 5100'F 3)
Safety injection delay 512 seconds (with offsite power) s 27 seconds (without offsite power) h)
Containment pressure Bounded - Based on minimum containment pressure of 14.7 psic 1)
Single failure Loss of one train j)
Control rod drop time N/A
'70'F is a statistical lower limit for the SI temperature based on actual plant data. Temperatures as low as the Tech Spec lower limit of 35'F are acceptable.
'I Al - 27
I Farley Pbot vo. Pmid 0.45 (0.28,0.43)
(0.36 0.43) i 0.4 o
ooo O
OS}8
~
c o 0 0 og" 0.35 o
(0,45,0.335)
CU o
h o o O
tw g
e DO 1
OoOF (0.28,0.315)
O e
S 0.25
'i>
(0..,5, 0.225) 0.2 0.25 0.3 0.35 0.4 0.45 0.5 l
PMID Figure 13-1. Farley Units 1/2 P110T/PMID Limits Superimposed on a Plot of All Possible Power Shapes for a Typical Fuel Cycle Al - 28 l
.m
NRC OunlifnNLl3 For the small break LOCA analysis, the submittal is not clear with regard to single failure assumptions and limiting conditions. %c submittal originally states that the limiting single f:.ilure is "that of an emergency power train." A few sentences down the submittal states that the "assumpt on of LOOP [ loss +f-offsite power] as the limiting single failure for small break LOCA i
is part of the NRC approved methodology." Please clearly state the assumed single failure, its affect on ECCS perfonnance and the bases for it being the limiting single failure for all cases.
Additionally, discuss the availability of offsite power in the analysis. State the basis for assuming the LOOP occurs coincident with the reactor trip. Why is this case more limiting than assuming offsite power is lost at or prior to the LOCA and why is this more limiting than assuming offsite power is available throughout the event? Describe the modeling of the reactor coolant pumps and explain why not mcxleiing the pump heat is acceptable. (WCAP 14723, Section 6 9)
SEC_Regense No 14 He NRC approved small break LOCA methodology is documented in % CAP 10054 P A.
" Westinghouse Small I!rcak ECCS Evaluation Model Using the NOTRUh1P Code." Section $
states that the appropriate inputs of Appendix K to 10 CFR 50 Part 46 are incorporated in the NOTRUMP small break LOCA ECCS evaluation model. This section discusacs the considerations which affect the severity of a small break LOCA transient, including considerations such as decay heat, availability of ofTsite power, and the worst single active failure. The section concludes that the NOTRUhtP small break LOCA ECCS evaluation model calculations performed with the Appendix K assumptions will clearly result in an overestimate of the severity of a small becak LOCA transient. It is Westinghouse's positian that the input assumptions documented in Section 5 have been established as a conservative input assuraption set as approved by the NRC.
Application of the input assumption set (including LOOP) as documented in Section 5 will result in a conservative small break LOCA analysis result for Farley. Furthennore, it is not necessary to evaluate rJternate single failure scenarios that produce minor PCT difTerences due to the overall conservatism contained within the approved input assumption set. The appropriateness of the Loss of OfTsite Power (LOOP) scenario with the corresponding limiting single failure of the loss of an emergency diesel has been previously documented in Reference i for the Parley units. Given the assumed LOOP at reactor trip, the limiting single failure is assumed to be the loss of an emergency diesci.
In SI3LOCA the LOOP is assumed to occur coincident with reactor trip, consistent with the generic studies documented in Reference 1. The time of reactor trip is typically very early in the transient (less than 14 seconds after break initiation for the limiting 't inch break in the uprate analysis), and the delay in LOOP from the beginning of the transient is considered to be inconsequential.
Neglecting pump heat in the SilLOCA analysis is the standard assumption of the approved application of NOTRUMP lbr Evaluation Model calculations as can be inferred from the response to RAI 440.18.0 and.140.28.0 given in Reference 2. This assumption is conservative for the followmg reasons. Until the time of pump trip, the pump heat would be 1 source of reduced subcooling in the broken loop cold leg His reduced subcooling would be translated directly into a reduction in the break flows for the approved Evaluation hkxlel subcookxl break flow model.
Al - 29
l l
1 Further, tMs reduction in subcooling could lead to an earlier loop scal clearing time which typically leads to reduced PCT.-
l Epiponse No.14 Rpfttracn 1.
WCAP 11145 P A," Westinghouse Small Break LOCA ECCS Evaluation Model Ocneric Study with the NOTRUMP Code," S. D. Rupprecht, et al., October 1986.
2.
Westinghouse letter SED SA 00449 from H. A. McIntyre (W) to P, H. Abramson (ANL),
July 26,1983.
I Whgm 11/14/97 i
NRC.Hucatiolthol I
f
' For the small break 1.OCA analysis, the submittal ind...tes that the ECCS flow is Jelivered to both the intact and broken loops "at RCS backpressure." Following a LOCA, the pressure in the broken loop will be lower than in the intact kops. As a result, the injected flow will preferentially go to the broken loop because it is at a lower pressure, Modeling both the intact and broken loops at RCS pressure will overpredict the flow going to the intact loops and underpredict the flow in the broken loops. Explain why your modeling assumptions acceptably bound plant conditions. Additionally, describe why your modeling approach climinates the need for the 150*F peak clad penalty. (WCAP 14723, Section 6 9/10) tiNC.Rupmse No I$
l in WCAP 14723 Section 6.1.2.3, the analysis description states that ECCS flow (i.e., high head injection)is delivered to both intr.ct and broken kops at RCS backpressure. His mean the local RCS pressure at the separate ECCS injection sites; i.e., multiple backpressures are explicitly modeled: Therefore, the marginal differences in pressure between the two modeled kops (i.e., one broken and two lumped intact kiops) are accounted for in the injection fiswrate calculation. This
. methodology is consistent with that which has been reviewed and approved in the SER to Reference 1. Explicit modeling of ECCS flow into the broken kop therefore obviates the need for the 150*F penalty.
Note, however, that the Parley uprate analysis, which was performed prior to final approval of Reference 1; only utilizes that portion of the Reference I methodology associated with ECCS flow mateling and does not take credit for the improved COSI condensation model, which would have hxl to significantly reduced PCTs.
Reponse No 15 Rtfttents 1.
'.WCAP 10054 P A, Addendum 2, Revision 1," Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety injection into the -
Broken loop and COSI Condensation Model," C. M. Thompson, et al., July 1997.
Whgm-l1/17/97 Al-30 m
NRC QunlintLNoJ6 The submittal indicates that a break of an injection line is less limiting than a break at the bottom of the cold leg. Please provide additional bases for this conclusion. Discuss the competing effects of degraded injection flow (decreasing ECCS performance) and the more rapid core depressurization (mcreasing ECCS performance) as a result of the break at the top of the cold leg. A quantitative assessment of the relative magnitude of these effects would be helpful.
(WCAP 14723, Section 6-9/10)
SNC Encens_Nol0 The conclusions provided in Reference I, are based on sensitivity runs which indicate that an analysis performed with the assumption that Si spills to containment back pressure would require the break to be located in the upper regions of the cold leg. lircats located high on the cold leg would result in lower PCTs due to improved capability to vent steam and therefore are less limiting, improved capability to vent steam lowers backpressure sooner in the intact loops, resulting in imptoved Si flow in the intact hiops. This tends to ofTset increased Si break flow in the broken k>op. Rcrer to Reference 1, particularly Section 5.0, for further discussion and generic quantitative assessment.
Enneme No. Ifiltfrancs 1.
WCAP 10054 P A, Addendum 2, Revision I," Addendum to the Westingimuse Small lireak ECCS Evaluation hiodel Using the NOTRUh1P Code: Safety injection into the Broken loop and COSI Condensation hiodel " C. hl. Thompson, et al., July 1997, Whpn - l 1/14/97 8EsCJhchtien Noll The submittal indicates that the reactor trip is modeled to occur at 1840 psia with a setpoint of 1865 psig; however, the uncertainty presented on pg 6 2 indicates that a A50 psi would be used.
Explain why the A50 uncertainty was not used for small break LOCA. (WCAP-14723. Section 610)
SNCEnpomdoJ2 The 50 psi uncertainty described on page 6 2 of the submittal is the uncertainty on the initial pressurizer pressure at steady state. This is conservatively taken as an adder to normal operating pressure of 2250 psia in the SilLOCA analysis leading to an assumed steady state pressure of 2300 psia. The reactor trip setpoint input for NOTRUh1P is the safety analysis limit that conservatively accounts for the setpoint pressure uncertainty. Sctpoint uncertainty calculations demonstrate that margin is maintained between the safety analysis limit (1825 psig) and the Technical Specifications reactor trip nominal setpoint (1865 psig).
.Whpn I1/14/97 Al - 31
i I
NRC_QuaticaNelB he submittal indicates that units "nny be subject to Si hafety injection) interruption or reduction while switching over to cold leg recirculation." Picase explain how the calculated core temperature is maintained at an acceptably low value if Cow can be interrupted for bo.h small and large breaks. Ilow is long-term cooling acceptability calculated and verified? OVCAF.
14723, Section 61!)
SECanponsc No.1H For SilLOCA, the Farley units do not experience a reduction or interruption of ECCS flow during the switchover to cold leg recirculation. Therefore, there is no impact on calculated core temperature or long-term core cooling. He description on page 6-11 is merely intended to provide the reviewer with a concise description of the evaluation of potential efTects of containment spray actuation during Sill OCA, whether applict.b'c to the Farley units or not. Note that the enthalpy increase portion of the issue does apply to the Farley units and was evaluated using plant specine information and found to have no impact on the results.
For 1 IILOCA, the high head ECCS injection How is continunus and the low head ECCS injection How is temporarily intermpted for less than four minutes during the switchover from injection to cold leg recirculation. The temporary inictruption oflow head ECCS flow casures the RilR pumps are not inadvertently damaged during switchover. This switchover sequence was included in the initial version of the Farley FS AR (Section 6.3). To quantify the potential impact of the temporary fow head FCCS How interruption would require a rew analysis. long term core cooling is addressed in Section 6.1.l.3 of WCAP-14723 (page 6-4).
W/rpn. I1/14/97 A SNC/mpe. I1/18/97 NRC.HuntienEoJS The rensitivity study for your analysis in('icates that the lower T.,, yicids more limiting results than a higher T,,, is there any physical reason you would espect that this would be the case for your plant design?
SNCJkmmKN;Ll2 For the Farley specine SilLOCA mtalyses, the PCT results are relatively insensitive to the analyzed temperature range. The ditTerence in results is a function of the model and the Farley-specine input assumptions, and it is not attributed to any physical phenomena.
For I.llLOCA, higher Tavg results in earlier Dashing throughout the primary side. The earlier degradation of reactor coolant pun p head due to two-phase conditions allows increased downllow through the core and increased blowdown cooling. The following Ogure rhows the PCT cfTect for the high and low Tavg cases. This reduction in claddinb temperature at the end of blowdown then carries through the renood portion of the transient, with the high Tavg case remaining cooler throughout.
W/rpn.11/1R/97 Al 32 t
a Revised in1tle! Transient
--~ ~ Hi gh f eve 2000 j
1800 --
^-
1000
- l
',sl,%
u NyY I
I
/
V m
,,f'.%
w
\\
Iy l t
i i
V I, 34ng e
l ',
)*51 qv w
ll3 Il 9'k#
.' ll 1.h o
w I
o 1200 n.
~
Ew
>==
1000 800 800 d
50 100 100 200 2!i0 300 Tim. (0)
Figure 191. PCT Comparison for Low T.,, and High T.,, for Farley Unit 2 Al.33
NRC_Questiodol0 Verify that the accident analyses assume a stuck tod following a reactor trip, the most limiting single failures, and a LOOP for each transient. Describe how uncertainties in the Fu and Fui are reficcted in the analysis SNC_RememdoJD The non LOCA and SI1LOCA safety analyses, in all cases, conservatively consider a stuck rod Ibilowing reactor trip and the limiting single failure. The LDLOCA analysis does not take credit for rod insertion; i.e., all rods are assumed to be fully withdrawn. 'ihe limiting single failures have been established for the Parley Nuclear Plant on an event specine basis and have been verified to be correct. All events that rely on ESF actuation are modeled assuming LOOP. In addition, the RCP locked rotor accident which does not require ESF actuation is also modeled assuming LOOP.
'the events are discussed in the Farley FS AR and the NSSS Liccasing Report (WCAP 14723).
For events that consider a loss of offsite power, the assumption is that the loss of ofTsite power occurs as a consequence of the unit trip wiuch causes instability on the power grid. The typical assumption is that the loss of ofTsite power, which results in a RCP wastdown, occurs 2 seconds following the turbine trip (which occurs as a result of reactor trip). *n his is a reasonable delay timo in that the turbine trip, generator trip, grid instability and loss of power to the RCP sequence will not occur instantaneously.
For non LOCA events, bounding analyses are perfonned for the various accidents in which and Fui ncluding uncertainties are used. On a reload-specine basis, Fn and Fut i
nmimum Fu including uncertainties are veriGed to meet the analyses assumptions.
Uncestainties in the Fu and Fuifor the Farley uprate BELOCA analysis are handled in the same manner as described in the approved Best Estimate LBLOCA Methodology (Reference 1). The uncertaintics in F and Fui, along with other power distribution attributes, are included in the 9
calculation of the power distribution uncertainty contribution to the overall uncertainty. The overall uncertainty is then calculated, resulting in the final estimate of the 95th percentile PCT for Farley Units I and 2 shown in Table 6.1.12 of Reference 2.
The SBLOCA analysis uses the Technical Speci6 cations values for Fu and Fui. The uncertainties in Fu and Fui are included in the Technical Speci0 cations values.
Refpome No. 20 RtfrIrnets 1.
Letter, R. C. Jones (USNRC) to N. J. Liparulo (W), " Acceptance for Referencing of the Topical Report WCAP 12945 (P), Westinghouse Code Qualification Document for Best Estimate loss-of Coolant Analysis," June 28,1996.
2.
WCAP 14723,"Farley Nuclear Plant Units I and 2 Power Uprate Project NSSS Licensing Report," January,1997.
Whgm. I1/lW Al-34
MRC.QuntiatNoll For a number of transients, the analy sis perfonned to support the overtemperature deha temperature and loverpowerj delta temperature setpoint license amendment change, the VANTAGE 5 fuellicense amendment change, or the existing analyses are credited. Please describe the modeling for these analyses and verify that a.li the changes associated with this l
license amendment request are correctly mateled. Modeling assumptions include the effects of th > in:rt ased power, reduced RCS Dow, reduced ECCS flow, reduced Tm, control rod optiraization and ZlRLO. If the analyses that are credited have not been submitted under a different approved license amendment request, please provide additional details regarding assumptions and approved codes. If the credited analyses do not model all the chaages associated with this submittal, provide a justi0 cation for not needing to explicitly model them.
SEC.Etspense No. 21 All of the mateling for the non LOCA and LOCA safety analyses either currently presented in the FSAR, or documented in conjunction with the uprate program, were performed using NRC appmved cales identined in the response to Question No. 8 above and have been verified to be cr rectly nuleled With respect to the plant uprate, all non LOCA analyses performed for the overtemperature delta temperature and overpower delta temperature setpoint license amendmcet were performed to bound both the current 1.lant conditions and operation at uprated conditions. The input assumptions, including reduced RCS flow, reduced ECCS flow, etc., were the same as those used in the transients reanalyzed for the oprate. Herefore, these analyses explicitly bound the proposed conditions for the uprate program.
He non LOCA analyses performed in support of the VANTAGE 5 licensing amendment were perfor r.u! at uprated power, llecausnome ana!ysis input assuncions difTered, the VANTAGES Chapti. 15 analyses which were not directly reanal> red for the plant uprate license amendment were reviewed with respect to all proposed uprate conditions and analysis input assumptions. Each of the VANTAGF 5 Chapter 15 analyses which maleled conditions related to the increased power N
level, redered RCS 110w and ZlRLO clad fuel have been confinned to support plant operation at the proposed uprate condithns. The only VANTAGE 5 Chapter 15 transient which credits the safety inject:an hystem and which was evaluated for the uprate is the feedline break event. For this transient, the redund ECS flow was explicitly modeled in the sensitivity calculation noted in response to Question No. 30, which follows, and therefore, has been appropriately considered. The main steamline break nnalysis, which credits actuation ot'the safety injection system, was reanalyzed and c.splicitly modeled the reduced ECCS flow, he Reactor Control System is modeled in analyses only when the response of the automatic rod control worsems *he consequences mhe transient. This system is not credited for event mitigation.
The Reactor Control System is modeled in the Chapter 15 analyses of the feedwnter malfunction (Section 15.2.10), execssive load increase (Section 15 2.1 l), and RCCA misalignment including dropped RCCA (Section ;5.2.3) events. Of these ennts, only the analysis of the dropped ",CCA transient is sensitive to ic Reactor Control System response, and therefore, only the conclusions of this analysis wrre potentially impacte : by the optimized setpoints. The dropped RCCA analysis supporting the uprate program exphcitly considers the efleet of the revised Reactor Control Sycem Al - M
l setpoints as noted in Section 6.2.3 of WCAP 14723 and, therefore, bounds the proposai uprate conditions. (Note, sensitivity calculations of the fealwater malfunction event, performed for the uprate program, also included the revised Reactor Control S) stem model and confinned that the anal) sis was not significantly impacted )
W/gls. I1/12/97 NBCSunt!0tlha 22
- the verbiage in chapter 6 of the topical report appears inconsistent with the in?oduction in section 6.2.0. For example, in section 6.2.0 the uncontrolled rod cluster control assembly withdrawal at power transient is listed as a transient that " Reanalysis perfonned in support of the FNP lFarley Nuclear Plant] uprate. ilowever, in sxtion 6.2.2 this transient is dispositioned by stating that NRC review and approval for this event was already received and no details ire included. To clarify, please provide a list of all the Chapter 15 transients; for cach, indicate if the analysis was performed to support the uprate, if analys s was perfonned earlier at the uprated conditions (including all other changes), or if analysis is not necessary with an explanation justifying the conclusion SEC.Enpetu9_No._22 in the example provided above of the uncontrolled RCCA withdrawal at power transient, this Section 6.2.0 item is annm nted to indicate that " Reanalysis performed in support of FNP uprate" o
and that "NRC review and approval received as part of OTAT/OPAT setpoint revisions." That is, the uncontrolled RCCA.ithdrawal analysis that was reviewed and approved for the OTAT/0 PAT setpoint revisions also included assumptions which bound the proposed uprate conditions. 'Ihis is also true for the uncontrolled boron dilution (Section 6.2.4 lcvaluatio1]), accidental depressurization of RCS (Section 6.2.12), and steam system piping failure at full power (Section 6.2.23) analyses.
The transient /crent summary in WCAP 14723 Section 6.2.0 accurately annotates (i.e., identMes) whether a new analysis was explicitly perfonned in support of the plant uprate conditions or an evaluation of the current FSAR analysis was performed. For most evaluation cases, the results demonstrated that current FSAR analyses would centinue to bound the conditions proposed through the license amendment. The exceptions are bric0y discussed below.
Start up of an ' nactive Reactor Coolant Loop (Section 6.2.6)
The evaluation of this transient demonstrates that this event is precluded by the plant Technical Speci6 cations. The event was initially analyzed for FNP when N 1 loop operation was considered as a future possibility, FNP has not pursued a license amendment for operation with a n: actor coolant kop out of service, and therefore, the transient was not re-analyzed at uprate conditions.
Excessive 11 cat Removal Due to Feedwater System Malfunction (Section 6.2.10)
The current FSAR analysis uas perfonnat for the VANTAGE 5 fuel upgrade. The analvsis conditions included the proposed uprated power level and corresponding RCS temperature and now conditions. A sensitivity calculation was performed for the limiting cases at ilFP and ilZP to Al - 36
i i
demonstrate that the analyses were not sensith c to the additional paramder changes included with i
the uprate license amendnwnt.
t With respect to dw postulated feedwater temperature redue;;on.ispect of this transient, a resised i
calculation was perforned to support an increase in the potential temperature reduction from 60'F j
to 65'F due to a postulated failure in tlw foodwater heaters.1bs calculation is discussed in I
Section 6.2.10.5 and is appropriately reficcted as an update to dw FSAR.
Excessive lead increase incident (Section 6.2.11)
The current FSAR analysis was performed for the VANTAGE $ fuel upgrade. The analysis conditions included the pn> posed uprated power level and corresponding RCS temperature ad f
flow conditions. An evaluation of this non limiting analysis was perfornwd by comparing the limiting transient conditions of peak power and temperature and minimum pressurizer pressure to de core thermal limits The Mastion demonstrated that the DNHR design basis is nwt.
- Major Rupture of a Main Feedwater Pipe (Section 6.2.20)
'Ihc current FSAR analysis was ;)ctformed for llw sclatively recent steam generator level tap relocation. Ihc analysis conditions included the proposed uprated power level and corresponding RCS conditions. A sensitivity calculation was performed for the limiting FSAR case in order to demonstrate that the FSAR analysis was not overly sensitive to the additional parameter changes included with the uprate license amendment. (The results of the evaluation are discussed in more detail in response to Question No. 30 below.)
w/gls.11/1297 NRC GuntionE9J2 1he submittal indicates that the optimized rod control system parameters (WCAP.14723, Section 6-62) are nxxicled in the accident analysis. Please describe these changes to the rod control system arxl how it n& cts the accident analysis.
SNC Response No. 23 Revised Reactor Control System setpoints and time constants listed on page 6 62 of WCAP 14723 include the following.
' I) ' Revised non-linear gain break points in the power mismatch channel. The bre:1 points were revised from 11% to 12% to minimize the efket of steady-state NIS fluctuations on the demanded rod motion.
- 2) 1hc Tavs filter time constant was increased from I second to 10 seconds to filter out signals associatal with RCS temperature spikes.
- 3) : The Tavg lead time constant was decreased from 50 seconds to 40 seconds to decrease the rate compensatio.: on signals associated with RCS temperature spikes.
Al-37'
, ~ -
t Together, these changes minini c undesired automatic rod stepping when the Reactor Control System is in autonatic mode.
Rod control is modeled in the safety analyses only when the a sponse of the automatic Reactor Control System worsens the consequences of the transient, T. :s system is not credited for event mitigation. The Reactor Controf System is modelcd in the Chapter 15 analyses of the fealwater malfunction (Section 15.2.10), excessive load increase (Section 15.2.11), and RCCA misaligmnent including drrpped RCCA (Section 15.2.3) cvents. Of these events, only the analysis of the dropped RCCA transient is sensitive to the Reactor Control System response; therefore, this analysis was perfonned to demonstrate the acceptability of the revised rod control parameters.
A dropped RCCA or RCCA bank event causes a rapid control bank withdrawal due to the large primary to secondary power mismatch. The control rods are subsequently reinserted as the automatic Reactor Control System attempts to restore the RCS average temperature to program conditions. The revised rod control parameters result in a slightly delayed bank reinsertion, and thus, a potential increase in the peak nuclear power. Ilowever, the analysis results demonstrated that the DNH design basis continues to be met.
As previously stated, the dropped RCCA analysis supporting the uprate program exphcitly considers the effect cf',he revised Reactor Control System parameters as noted in Section 6.2.3 of WCAP 14723, and therefore, bounds the proposed conditioas. Sensitivity calculations of the feedwater nulfunction eve '
rformed for the uprate prograa, clso included the revised Reactor Control System model a
.mned that the analysis was not significantly impacted.
W/gls 11/12/97 NILC_Qttatiar1Eo2 On page 6 62 of WCAP 14723, the plant changes modeled in the accident analysis are described.
1he reduction in the required RCS flow is not included in this discussion. Please verify that the reduction in RCS flow is modeled in all the accident analyses credited in the submittal.
SNC Respmuthp]
As discussed in WCAP 14723 Section 6.0.1 and as noted in WCAP-14723 Table 6.2.01 (page 6-65), the appropriate value of RCS flow, including uncertainties, was properly modeled in the uprate accider,t analyses.
Whgm 11/l.1/97 NRC_Qtintinn No. 2$
It is appropriate to use the Revised 1hermal Design Procedure when evaluating the critical heat flux or departure from nucleate boiling (DNH) and then use nominal values for oower, pressure, and temperature. Ilowever, the uncertainties should be included in evaluating c..e other limits
. such as peak pressure and linear heat rate. For the transients that challenge the limits other th.v Al-38
i DNil and the revised thermal design procedure was used, please evaluate the need to perfonn an analysis to verify that the other limits are not exemled.
SNC Refpenstle_21 Transients which are analyzed using the Revised Thennal Design Procedure (RTDP) to address the DNil consequences, and which also challenge non DNil analysis limits, are discussed below. In cach case, the non DNil aspects have been appropriately analyzed considering initial condition uncertainties.
The loss ofload/ turbine trip event clearly challenges the RCS overpressurization criteria wherein the peak calculated pressure is 1.5 psia from the analysis limit. The analysis of this condition conservatively models initial condition uncertainties for power, RCS average temperature and pressurizer pressure, including thermal design flow, as indicated in Section 6.2.7.
Similarly, initial condition uncertainties are modeled in the analysis of the locked rotor event (Section 6.2.21) for the peak RCS pressure and clad temperature transients, and in the analysis of the inadvertent operation of ECCS during power operation (Section 6.2.14) for the pressurizer filling transient.
In addition to demonstrating that the DNil design basis is met, the steamline break at power analysis (Section 6.2.23) challenges the peak linear heat rate limit (kW/ A). This analysis, previously reviewed for the overtemperature delta temperature and overpower delta temperature setpoint program, did not model initial condition uncertainties. The potential impact of explicitly modeling uncertainties in this analyses was recognized, and it was resolved by identifying existing conservatism in the analysis method which offsets the potential impact for the FNP uprate. Note that the steamline break analysis (along with other limiting transients) is specifically reviewed as part of the reload process, and thereby, confirms that the analysis criteria continue to be met for cach cycle.
W/gls. I1/12/97 NBC_QmlimtFA26 For the loss of electric load analysis, what was assumed for the accumulation in the safety relief valves? What are the peak primary and secontry pressures calculated?
SNC lMppnse No. 26 The loss ofload/ turbine trip (LOl/fT) analysis explicitly models the pressurizer loop seals with a r,rge time of 1.6 seconds. The loop seals are assumed to begin to purge when the pressuriaer pressure excmis the opening pressure of the safety valves. No steam reliefis credited until after the loop seal is fully purged (i e.,1.6 seconds), llecause of the rapid RCS pressurization during this portiv, of the transient, the pressurizer pressure is well above the fully accumulated pressure (3% above opening pressure) of the valv:s. Once the laop seals are purged, the valves are assumed to setieve at up to their full capacity, A 3% accumulation pressure is, therefore, enveloped by the analysis model. The peak RCS pressure (at the !(CP outlet) calculated in the Chapter 15 analysis was 2747 psia compared to a limit value of 27414.5 psia.
Al-39
There are five main steam safety valves on each steamline with one valve assigned to each bank associated with the three steamlines. In al; non LOCA safety analyses,3% accumulation is modeled for banks I through 3,2% accumulation is modeled for bank 4, and 10 psi accumulation is modeled for bank 5. Thcse assumptions are supported by test data on the Farley plant main steam safety valves. 'Ihc peak nuin steam system pressure calculated in the Chapter IS LOI flT analysis was i199 psia compaced to a limit of 1208.5 psia.
ERC_QuesliettNo. 26 SuppkmendBsfttence November in.1997 NRC/SNC Conferenec_ Cal])
1he reactor trip function response times assumed in the ren LOCA FSAR analyses for the FNP uprate are identified on page C-63 of WCAP-14723. The data indicates a change to the response time for the high pressurir.er pressure reactor trip function from 2 seconds to I second. Identify the reason for the change and justification.
SEC3cfpgn3sNLifduppkment As indicated by the response to Question No. 26, there is little margin between the peak RCS pressure and the pressure limit in the locs ofload/ turbine trip analysis. The reduced response time assumption was credited in order to generate margin in the analysis, and thereby, demonstrate that the analysis acceptance criterion for RCS pressure continues to be met. The response time testing history for this trip function was reviewed in support of this chan:. The review confirmed that the Reactor Trip System consistently responded within the revised I second requirement for pressurizcr high pressure. In support of the FNP uprate, the I second requirement for this functiaa will be incorporated into tia piant response time testing procedures and FSA R.
W/gls. l1/lR/97 ERC_QuCElleAELil For the mulysis of the loss of nonemergency ac power to plant auxiliaries analysis please describe the sequence of events in greater detail Describe why the loss of flow is a more limiting DNH event and explain why the assumptions are chosen to assure limiting results. Additionally, describe th initisl conditions relating to the important parameters.
SECRCSP2nE9.E0.11 The loss of non<mergency ac power eve..t. as presented in the FNP FSAR, demonstrates that with a loss of o(Tsite power that the auxiliary feedwater (AFW) system is capable of removing the residual heat generated in the RCS. The analysis under this section is efTectively a loss of nornal feedwater event, with a subsequent loss of ofTsite power. The net efTect is that the reactor coolant pumps coast down, fbliowing turbine trip, and that the emergency diesel generators are required to power the motor-driven AFW pumps. For this reason, the sequence of events and results presented are similar to those presented for the loss of normal feedwater event, The assumed initial conditions are the same as those presented in WCAP 14723 Section 6.2.8.2 (loss of nornal feedwater), except as noted in Section 6.2.9.2 (loss of non-cmergency ac power).
Al - 40
A loss of offsite power explicitly analyzal as an initiating event would result in an inunodiate coastdown of the RCPs, similar to a loss of voltage to (Se RCP power supply busses. This event would be less limiting than the complete loss of flow event analyzed in Section 6.2.17 since an immediate reactor trip would also occur due to RCP undervoltage. Therefore, the DNB consequences of a loss ornon-cmergency.,e power event for the FNP are bounded by the analysis of the complete loss of flow event in which the reactor trip is not assumed to occur until generated by a low reactor coolant flow trip signal.
W/gls 11/12/97 NRC_Qmtion No 28 For the inadvertent ECCS actuation, the description in the submittal is not consistent with regard to the credit given the power-operated relief valve (PORV) to actuate on demand. Page 6124 states,"PORVs are not assumed as an automatic pressure control function for the pressurizer filling case." llowever, later discussion states, "PORV avaMility must be assural by manually opening the block valve to allow the associated PORV to actuate on demand." Please clarify the situation by describing the reliance of the PORVs. For the overfdling case, please indicate if the acceptance criteria (pressurizer is permitted to go water solid), the operator actions (7 minutes to open the block valve), and reliance on the actuation of the PORY have been approved by the NRC for this transient. Desca;be the single failure assumptions.
SNCJmnense No. 28 1he analysis of the inadvertent ECCS actuation at power event presented in WCAP 14723 t:.kes credit for actuat!on of at least one PORV following operator action to ensure that onc PORV is available. The normal operating aligrunent is with the PORV block valves open and the PORVs in automatic mode. Nevertheless, the analysis assumes that both block valves are closed and the PORVs are in manual mode, and therefore, the analysis models the operator manually opening one PORV block valve and subsequently opening the respective PORV, if required. The statement that the POhV is not assumed as an automatic pressure control function is intended to distinguish between the valve actuating immediately when the setpoint is exceeded, versus waiting until the operator has taken any neerled action.
The safety analysis predicts that the pressurizer may go water solid, and hence, takn credit for water solid discharge through the PORV and associated discharge piping. The analysis assumes that operator action to establish a path for PORV water relief occurs within 7 minutes. This analysis assumption is consistent with Farley emergency response procedure (i.e., E-0) instructions, which direct the operator to ensure that at least one PORV block valve is open and that the associated PORV is available in reponse to a pressurizer overfill / overpressure transient.
While a part of the FNP FSAR Chapter 15 analysis, reliance on actuation of the PORV for water n:liefin this manner has not been explicitly reviewed and approved by the NRC. A dditionally, the analysis does not consider a sing,1c failure with respect to the PORV actuatica since this action is associatal with the operator action.
W/gls 11/12/97 & SNC/mge 11/15/97 Al-41
-. ~ _ - - - - -. -. -
. ~.
h MRC.QuntiD!tNo. 29 Please provide inore information regarding the analysis of the main steamline break (MSLB).
. Please provide an explanation why _the analyses performed, with the assumed single failures, are f
bounding (with assumptions made regard;ng rod motion). Discuss the LOOP assumptions and justify that they are bounding. %c conclusions for the MSLB only state that the DNB design basis is met. Please state the design basis and prmide the results. W8:at is 1 :e minimum D~ f!J i
5 r
and wlmt percentage of the rods experience DNB (if applicable)? Verify that the primary and secondary pressaic is maintained below acceph:-
?. sign limits, considering the potential far brittle and ductile failurcs. He response to quw on 8 references WCAP 9226, Rev.1. " Reactor Core Response to Excessive Secondary Steam Releases," tojustify the use of a conscivatively low flow rather than the use of a conservatively high flow. Please provide a copy of this topical (or provide a refercacc ifit has been docketed) and justify that the conclusion is applicable to your plant. If maximum vs. minimum RCS flows are used, would the cooldown be more severe and cause a violation of the cooldown limits associated with brittle or ductile failures? Does the flow area through the main steamline nozzle assumed in the analysis account for or bound the design tolerances and themm1 expansion of the metal?
SNC Remonse No 29 As previously noted, the MSLB analysis was perfonned tollowing the methodology presented in WCAP 9226 Revision 1. NkC spproval of the methodology was provided in:
Letter from / shok C. Thadani (Assistant Director for Systems, Division of Engineering &
Systeins Technology, Office of Nuclear Reactor Regulation, USNRC) to W. J. Johnson (Manager, Nuclear Safety Depaament, Westinghouse Electric Corporation), " Acceptance for Reference of Licensing Topical Report, WCAP 9226.P/9227 NP, Reactor Core Response to Excessive Secondary Steam Releases," January 31,1989.
%c MSLB topical report discusses analysis assumptions and models related to the limiting single failure, rod motion, stuck rod conditions, LOOP, ctc. A copy of the letter and WCAP can be provided if not retrievabic unar the above reference.
%c minimum DNilR for the main steamline break event was shown to be greater than the design basis limit of i A5; therefore, the DNB design basis is met. It should be noted that as part of the reload safety anal e, the steamline break analysis is evaluated, and a cycle specific minimum s
DNDR is calculated. As such, the intent of the uprate analysis was to demonstrate that the design basis can be met with an uprated core design. %c cycle-specific analysis will determine the final margin to the DNBR limit of 1 A5.
The peak RCS and main steam system pressures do not exceed their initial values during the analyzed transient, and therefore, do not challenge their respective pressure limits.
The base plant desigin in the topical report include a 3 loop plant with a 12 foot core, an NSSS power level of 2785 MWt, and model 51 steam generators, among other similarities to the uprated Farley plant configuration. Therefore, sensitivity analyses such as thatjustifying use of a
- conservatively low flow (WCAP-9226 Section 3.1.1.8), may be directly correlated to the Farley plant.
- Al - 42
i
%c acumed maximum break sia is based on the nominal flow area of the flow nozzle integral to the steam generator. The area is not increased to account for design tolerances or thermal expansion of the metal. Ilowever, the break flow h conservatively maximized by assuming the Moody correlation with fl/d = 0 (noted in WCAP 9226 Section 2.2.2.6 and Section 3.1,l.2) and conservative moisture carryover assumptions (WCAP-9226 Section 3.1.1.9). Therefore, the overall break area and blowdon model are considered conservative for the transient.
The Chapter 15 steamline break analysis does not provide the basis for the evaluation of the potential for brittle and ductile failures, in fact, the thick metal masses in contact with the coolant, which would tend to help mitigate the cooldown of the RCS, are ignored during the evaluation of the transient.
Wigls & jsk 11/.2/97 NRC Ouestion No. 30 The analysis results are not presented in the submittal or the FSAR for the main feedline rupture.
Please identify the acceptance criteria and present the results of the analysis. Is fuel failure l predicted), what is the lowest fiNBR, arc Part 100 limits met? The FS AR analysis is referenced for this event. Please describe how the changes with this amendment are modeled in the FS AR analysis and justify why the results continue to be bounding. Discuss the LOOP assumptions and justify that they are t>ounding.
SNC Respans No 30 The eu rent licensing basis main feedline break analysis, presented in the Farley Nuclear Plant FS AR and referenced in the uprate licensing submittal, was performed for the steam generator level tap relocation license amendment. This analysis was based on the same core power level and general plant conditbr.s as proposed for the FNP uprate, and therefore, was evaluated in support of the uprate licensing submittal.
The icedline hO evaluation was performed via a sensitivity calculation of the limiting FSAR case. The sciain calculation quantified the impact of changes in secondary parameters relatea to the plant uprate assumptions compared to current plant parameter assumptions. The calculation Jemonstrated that there was no significant change in the RCS and main steam system pressure I
transients or peak pressure values while a small reduction in the minimum margin to hot leg sat uw. t im approximately 34 F to 28"F (compared to a limit value of 0*F), was calculated.
Theure
..ie fety analysis criteria were not challenged as concluded in Section 6.2.20 of WCAP-11723 (page 6-176).
To conservatively assure that the core remains in place and geometrically intact with no loss of core cooling capability, the analysis demonstrates that no bulk boiling occurs in the primary coolant system following a feedline rupture prior to the time that heat terr. oval capability of the steam generators being fed auxiliary feedwater exceeds the core heat generation. His criterion is more linr'ing and is adopted for convenience. The minimum DNDR is not explicitly calculated for this trar sent, nor is this event limiting with respect to DNB. With respect to activity releases, the linuting consequences of postulated secondary side ruptures result from the steamline break event.
Al - 43
With regard to LOOP assumptions, the original Farley feedline break analysis presented in the initial edition of the FNP FSAR assumed a loss of offsite power after the reactor trip. His assumption results in a loss of forced RCS flow and reliance on natural circulation for decay heat removal. His assumption is explicity modeled in die current Farley analysis.
W/gle 11/1R/97 NRC Ouestion No. 31 He locked rotor event assumes a " conservatively largt s osolute value) of the Doppler-only power cocmcient." Please justify this assumption. A iarge absolute Doppler pon :r coemeient would appear to be non-conservative in a heat up/over power event.
SNC Rs_sponse No. 31 Sensitivity studies used to develop the analysis method of the locked rotor event have shown that a large (absolute value) value of the Doppler-only power coemeient is conservative. He locked rotor event, as analyzed for Farley, results in a reactor trip signal (on low RCS loop flow) and initial rod motion in less than 1,5 seconds. De limiting transient conditions also occur very quickly (i.e., maximum rods in-DNB, RCS pressure, and PCT occur within 4 seconds). Herefore, the time of rod motion and subsequent nuclear power transient are key to the analysis. He dominant effect of the Doppler-only coef0cient occurs during rod motion while nuclear power is less than nominal and rapidly decreasing. Hence, the assumed large (absolute value) of Doppler-only power coef6cient is conservative since it results in a greater positive reactivity insertion during trip, and thus slows fne power decrease.
W/gk. I 1/12/97 NRC Ourstion No. 32 For the rod ejection accident, provide the results of the analysis (percentage of rods experiencing DNB, percentage of rods experiencing fuel melt, peak pressure). The analysis description in the submittal is not sumcient to determine if the analysis techniques are in conformance with the approved methodology, Verify that the analysis inputs and analysis technique was performed in compliance with Regulatory Guide 1.77, " Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors," and the approved methodology, WCAP-7588-1-A. Any deviations should bejustified.
SNC Response No. 32 De rod ejection analysis results, with respect to the percentage of rods which experience DNB and fuel melt, are used as input to the radiological analysis. In each case, the assumptions used in the
- radiological analysis are based on the analysis and methods presented in WCAP-7588 Revision I-A, which also supports the current plant licensing basis. The results of the WCAP analysis conclude that that no more than 10% of the fuci rods experience DNB following a rod ejection event, and that fuel centerline melt is only reached within fuel rods which enter DNB.
AI - 44
~-
5
- 7 Taking into account the radial power distribution in the ejected rod condition, this results in less -
than 5% of the fuel rods experiencing centerhnamciting:-Of the rodmperiencing melt;vnhiO% -
of the radial fucipe!!ct areawii!M% of the axial length will experience melting, thus the total melt
.,. ~;-voliihiiiiic~ ss than 0.25%, Note that for Farley, the analysis showed that a maximum of only 3.09% (<<10%) fuel melting occurred at the centerline in the hottest rod; thus, the number of
- affected reds and total fuel melt reported above is conservative.
' He overpressure analysis supporting the Farley uprate as well as the current licensing basis, was also based on the analysis and methoda docume.tA in the topical report. From Sections 2.6 and 4.4 of WCAP-7588, the pressure surge is caiculated using conventional heat transfer from the fuel and prompt heat generation in the coolant, WCAP 7588 Section 4.4 documents the results of i
a limiting RCS overpressure transient resulting from a rod ejection event at beginning-of life conditions. A conservative ejected rod worth, approximetely 3 times the ejected rod worth -
assumed in the FNP FSAR analysis for core considerations, is assumed with the core power generation weighted to the hot quadrant. This maximizes voiding in the hot channel, while no credit was taken for a reduced heat flux due to DNB. As stated in the FSAR (Section 15,4.6.2.3.6), the peak RCS pressure does not exceed that which would cause reactor pressure vessel stress to exceed the faulted condition stress limits. This overpressure calculation continues to provide a conservative analysis for FNP.
%cre were no deviations from :he approved analysis method in the analysis supporting the Farley plant uprate.
Radiological assumptions for evaluation of the consequences of a control rod ejection are consistent with those of Regulatory Guide 1.77, Appendix B with the following clarifications.
Ic.
%e fraction ofiodines assumed to be in the fuel-clad gap is assumed to be 12%.
2d.
Dose conversion factors from ICRP 30 are used.
2c, Doses are calculated with the TACTS computer program.
l 2g.
Site specific atmospheric diffusion parameters (x/Q) described in FSAR section 2.3 are used.
W/gls. I1/12/97 A SCS/ jaw. I1/14/97 4
- NRC Ouestion No. 33 For the steam genercor tube rupture analysis, please provide a summary of the analysis with the calculational results. What are the assumption and limiting sing!c failures? Provide an analysis for the steam generator overfill calculations with the analysis, assumptions, and conclusions presented.
i SNC Resnonse No. 33 s
WCAP 14723, Section 6.3, contains the Thermal / Hydraulic (T/II) results for the SGTR. These results provided input to the dose analysis provided in the BOP Licensing Report. Section 6.3 also z lists the pertinent analysis assumptions for the T/H portion of the analysis. The minimum auxiliary feedwater was assumed,. which is consistent with a single failure in the auxiliary feedwater system..
Al-45
~ -
i
__;,_,__._; Radioloscal.as>.umptions.for minationaf 41a. steam gerareter teW1upium were ptciideTin~5NC --
letter to NRC dated August 5,1927 'm Question No. 2, Table F, page 36), and the dose consequences were provided in the BOP Licensing Report.
_ Whgm A SCS/ jaw.11/14/97 l
l
-s
~l I
Al-46
)
l t
ATTACllMENT II -
SNC Response to NRC Request for AdditionalInformation Related To Power Uprate Submittal-Joseph M. Farley Nuclear Plant, Units 1 & 2 -
SNC RESPONSE TO NRC QUESTION NO.12 Westinghouse Electric Corporation letter CAW-97-Il88 dated November 18,1997, with accompanying affidavit, preprietary information notice, and copyright notice.
NSD-sal ESI-97-645, Proprietary Class 2C, "SNC Response to NRC RAI on DELOCA."
NSD sal-ESI-97-647, Non Proprietary Class 3,"SNC Response to NRC RAI on DELOCA."
!