Text

Rev 1 to Supplemental Reload Licensing Submittal for Pilgrim Nuclear Power Station Reload 7 Cycle 8
	ML20062E365
Person / Time
Site:	Pilgrim
Issue date:	09/30/1990
From:	Charnley J, Hansen E; GENERAL ELECTRIC CO.
To:	;
Shared Package
ML20062E357	List: BECO-90-136, Application for Amend to License DPR-35,revising Tech Specs to Upgrade Safety Limit Min Critical Power Ratio & Revise Operating Limit Min Critical Power Ratio.Revs Will Maximize Fuel Utilization During Fuel Cycle; ML20062E361; ML20062E365;
References
	23A4800, 23A4800-R01, 23A4800-R1, NUDOCS 9011200140
	Download: ML20062E365 (16)
	v • d • e

{{#Wiki_filter:. c i GE Nuclear Energ,- 23A4800 I Ke51sion 1 Class I September 1990 23A4800, Rev.1 - Supplemental Reload Licensing Submittal } I for - Pilgrim Nuclear Power Station l Reloai4_7 Cycle 8: i i .i s .j f:f.i N E.d (a\\%\\,[' Approved: / . Approved: nsen, - Manager j/J.S. ha ey, a ager Fuel Lleg Reload Nuclear Engineering i .t ,i ( l j l' 9011200140 901108 h PDR ADOCK 05000293 ~i P PDC pj l .6

y i Pilgrim. zum - Reload 7 Rev 1-l4 Important Notice llegarding Contents 'of This Report Please read carefully i This report was prepared by General Electric Company (GE) solely for Boston Edison Company (BECo) for BECo's use with the U. S Nticlear Regulatory Commission (USNRC) for f amending BEco's operating license of the Pilgrim N'idur Power Station, The information' contained in this report is believed by GE to be an accurate and true representation of the facts known, obtained or provided to GE at the time this report was prepared. The only undertakings of GE respecting information in this document are contained in th'e - contract between BECo and GE for fuel bundle fabrication and related services for Pilgrim' q Nuclear Power Station, and nothing contained in this document shall be construed as changing said contract. The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither GE nor any of the contributors to this document makes any representation or warranty j (expressed or implied) as to the completeness, accuracy or usefulness of the information' contained - in this document or that such use of such information may not infringe privately owned rights; nor j do they assume any responsibility for liability or damage of any kind which may result from such I use of such information. a y i ~k

]

R . t Page 2 i. dl J .,s ,........ ~.. _.)

y I m4sm Pilgrim. Ret 1-i Reload 7 h > Acknowledgmenti a-The engineering and reload licensing analyses which form the' technical basis of this Supple- ,} mental Reload Licensing Submittal,'were performed by E. O. Thacker II of the Fuel Engineering j Section. The Supplemental Reload Licensing Submittal was prepared by P A. Lambert and verified by O. O. Jones of Regulatory and Analysis Services. 3 'k i i 'I I c d i L l L 1 l . Page 3 - .i a

i J nam. 1 2 Pilgrim Rev 1 I Reload 7 ll 1.- Plant unique Items (1,0)* ' r Appendix A: Increased Core Flow Throughout the Cycle : 2. - Reload Fuel Bundles (1.0 and 2.0) I Fuel Tvos_ Cvele Imaded Number 1 Irradlated P8DRB282 5 24 + P8DRB265H 6 60 j l P8DRB282 6 112 BPSDRB282 7 -32 P8DRB282 7 160 t New BP8DRB300 8 192-Total 580-3. Reference Core Loading Pattern (3.2.1) mwd /MT ' mwd /ST Nominal previous cycle core average exposure at end of-cycle: _18,362 16,658 L Minimum previous cycle core average exposure at end of cycle ..18,362-16,658 .from cold shutdown considerations: 9 Assumed reload cycle core average exposure at end-of cycle: 21,018 19,067 Core loading pattern: Figure 1 'l \\ 1 1

( -) refers to area of discussion in General Electric Standhrd Application for Reactor Fuel, NEDE 24011 P A 9, September 1988; a letter "S" preceding the number refers to the U.S. Sup-plement, NEDE 24011 P A 9 C$, September 1988.

Page 4 ' 1

i e-2.w sm Pilgrim Rev 1 Reload 7 Cyculated Core Effective Multiplication and Control Systeni Worth No Volds; } 4. 20 C (3.2.4.1 and 3.2.4.2) } Beginning of Cycle, K,,, i 1.116- -l Uncontrolied EJ1 0.962 Fully controlled a 0.990 Strongest control rod out R, Maximum increase in cold core reactivity with exposure 0.000 into cycle, AK 5. Standby Liquid Control System Shutdown Capability (3.2.4.3) Boron Shutdown Margin (AK) (IED) (20'C. Xenon Free) l 675 0.040 6. Reload Unique GETAE AOO Analysis initial Condition Parameters (S.2.3) BOC8 to BOC8+8767 mwd /MT(BOC8+7953' mwd /ST) 'f Exposure: Fuel Peakinc Factors Bundle Power Bundle Flow Initial L Design local Radial Mia] R Factor (MWt) _ (1.000 lb/hr) MCPR b l BP/P8x8R 1.20 1.62 1.40 1.051 5.454 108.6-1.44 Exposure: BOC8 + 8767 mwd /MT (BOC8 + 7953 mwd /ST) to EOC8 i !~ L Bundle Power ' Bundle Flow - Initial-l Fuel Peakinc Factors Design Lgsal Radial A3ial R Factor-(MWt1 - (1.000 lb/hr) MCPR ( l BP/P8x8R 1.20 1.59 1.40 1.051 5355 ! 109.2 1.47: . 3 7. Selected Margin Improvement Options (S.5.1)- I Recirculation pump trip: No l l . Rod withdrawallimiter: No Thermal power monitor: No Improved scram time: No 1 Exposure dependent limits: Yes. Exposure points analyzed: 2 Page 5 -l 4

f s. nus*- Pilgrim ' Rev 1 Reload 7 8. Operating Flexibility Options (S.5.2) I Single loop operation: Yes Load line limit: Yes Extended load line limit: Yes i Increased core flow: Yes Flow point analyzed: 107.5 % Feedwater temperature reduction: No ARTS Program: No Maximum extended operating domain: No-1 9. Core wide AOO Analysis Results (S.2.2) Methods used: GENESIS Nuclear with G3 MINI ODYN and GEXL ACPR Flux O/A-l Event (% NBR) (% NBR) -

BP/P8x8R Figure Exposure: BOC8 to BOC8+8767 mwd /MT (BOC8+7953 mwd /ST).

Load rejection without 651 125-032 2 bypass Feedwater controller 663' 132 037 3 . failure Exposure: BOC8+8767 mwd /MT (BOC8+7953 mwd /ST) to EOC8 1 Load rejection without 730 126 035 4 l bypass L Feedwater controller 718 134: 0.40 5 L, . failure 3 l l L Page 6 g t ,+ ,-n

n i a: mam Pilgrim Rev. 1 - Reload 7 t

10. Local Rod Withdrawal' Error (With_ Limiting Instrument Failure) AOO Summary l

] 4 (S.2 2.1.5) q Generic Bounding Analysis Results - 4 4 Rod Block ACPR Readine BP/P8x8R 104 0.13-105 0.16 i 106 0.19 107 0.22 ~ 108 0.28 109 032 110 036 Setpoint selected: 110

11. Cycle MCPR Values (4.3.1 and S.2.2)

Safetylimit: 1.04 Non-nressurization events: Exposure range: BOC8 to EOC8 BP/P8x8R i Loss of feedwater heating 1.21* Fuelloading error 1.23* I Rod withdrawal error - .1.40 t Pressurization events: Ontion A Ont on B ' BP/P8x8R D.PJM Exposure range: BOC8 to BOCI,+8767 mwd /MT (BOC8+7953 mwd /ST) q Load rejection without 1A2-bypass Feedwater controller 1.48 a j., failure

The minimum MCPR value required by the ECCS analysis is 1.24.

" Option B is not available for this exposure.- Page 7 l I o

1 4 Pilgrim mam y Rev 1 Erlpad 7

11. Cycle MCPR Values (43.1 and S.2.2)' (continued)<

s Pressurization events: i Ontion A Ootion B l BP/P8x8R BP/P8x8R Expecure range: BOC8+8767 mwd /MT (BOC8+7953 mwd /ST) to EOC8 Load rejection without 1.45 1.40 bypass Feedwater controller 1.51 1.41 failure 4 (

12. Overpressurization Analysis Summary (S.3)

Pd Pv - I Event (pig) (psg) Plant Resnonse MSIV closure (Oux scram) 1317 1333-Figure 6 9 4

13. Loading Error Results (S.2.2.3.7)

Variable water gap misoriented bundle analysis: Yes Event ACPR Misoriented fuel bundle 0.19 4 l

14. Control Rod Drop Analysis Results (S.2.23.1)

Pilgrim Nuclear Power Station is a banked position withdrawal sequence plant; therefore the - control rod drop. analysis is not required, nkC approval is documented in NEDE 240ll P A US, September 1988. -I i i Page 8 3

e 1 Pilgrim 2msw Rev 1 Reload */ i t

15. Stability Analysis Results (S.2.4) 1 y

.p t Pilgrim Nuclear Power Station is exempt from the current requirement to submit a " cycle specific stability analysis as documented in the letter, C. O. Thomas (NRC) to H. C., Pfefierlen (GE), Acceptance for Referencing of Licerning Topical Report NEDE 24011 Rev. 6, ' ~ Amendment 8, 'Thennal Hydraulic Stability Amendment to GESTAR 11, ' April 24,1985. i k Pilgrim Nuclear Power Station recognizes the issuance of NRC Bulletin No. 88 07, i Supplement 1, Power Oscillations in Bol"*1g H'ater Reactors (BWRs), and will comply with the f 1 ~ recommendations contained therein'.

16. Loss of coolant Accident Results (S.2.5.2) j LOCA method used: SAFE /REFLOOD/ CHASTE -

See Loss of Coolant Accident Analysis Report for Pilgrim Nuclear Power Station, NEDO 21696, August 1977 (as amended). ,y ? t 'i I 1j I 1., Page 9 3

4 zwse 4 - . Pilgrim Rev 1 .l Reload 7 - MEMMMMM MMMMMMMMM MMMMMMMMMM4

MMMMMMMMMMMEs4

MMMMMMMMMMMMM

MMMMMMMMMMMME

M M M M M M M M M M M M E

':: M M M M M M M M M M M M M '::MMMMMMMMMMMME

M M M M M M is M M M M M E MEMMMMMMMER MMMEsMMMER L

MMMMMMM Il1IIIIl.lIlI11 4 1 3 5 7:9111315171921232527293133353739414345474951 . FUEL TYPE l o 1 A = P80RB282. D = P8DRB282-B = P8DRB265H' E = BP80RB282, C = P80RB282 'F = BP80RB300: 4 i Figure 1 Reference Core IA,ading Pattern l Page 10 '

' nus* > 6 Pilgrim Rew 1 Reload 7 YttW

he r

m. [ L /~ y=.4 / =. g 7 >y 1 D% =.

3

._..E s. u 4. y. tid (uC305) tid (ECac5), i f - jf EP kT',i,' [s

- ')

(\\ y ! 't 5 minn c =. ,_v i i 5 ]'s uaf A I "" -~ Wu i'- E --J

=,,,

n E 3, #, , u, =%. is. tid (utoes) - tid (ucoes) - r-4- ' Figure 2 Plant Response to Imad Rejection without Bvpass (BOCS to BOC8+8767 hnVd/MT [BOCS+7953 mwd /ST)) Pop 11 i -v +y

a 'PL' grim - zw. o Rev i-Reload 7 E ? t f 3.k (.M! Eitt V IA H[ Al f(V1 N1 # i_. 1 \\ i ~ '4 o m ::, =m - TM (ECec5) Tid (MC605) !M W

W u

/ . m

.. yp

[... 1 = = :- c 5 l i e_. /u i I \\ l i J or ind (ucacsi' tid (ucoesr .1 \\ 8 - Figure 3' Plant Response to Feedwater Controller Failure :. ,1 -(DOC 8 to BCC84 0767 mwd /MT [BOC8+7953 mwd /ST)) Page 12 .j 1' 1

q .1 2wsm u 8-Pilgrim ? Rev 1 - Reload 7 .t f J !E9dkw '@iipi@"" n f ^ d, +i.N ,/ ^ e,.. l i \\ N~ 3 .[;. 2. . #. = = t x (uc cs) l1 ' iw <scocu. t !goio ag$b!b!I g g u.s.at) f i evm !yygg%p)b i g NNY hif fddv / ~ i. j i =. v' .f .,A ^ 9; y . )'^.v ), =.m oc \\f*v-: =i G ( t f g 1 h: :. 8 c l-k 'i .= m. m. 1w (ut.cs) Tw (we cs) I \\ 1 i l i Figure 4 Plant Response to Generator Land Rejection without Bypass (BOC8+8767 mwd /MT [BOC8+7953 mwd /ST) to EOC8) 3 i Pase u j l i . +

8' Pilgrim - n^am Rev I-l Reload 7 .J . a ~t $#f a ,y al r t y, g

wv],h,d(F51) 1 L MI!

=j ) 1RON rgd i N 4 v l l i l,i-. - iac.e e a g '" ' w, \\ a 'l t

1

,.4 -- 4 i) ..m .m sg i tid (ECX$) tid (WCMS) - i f e $l0$F41) 1 VQ1 $ A*lthibt' IL y (IN* ., hh ' . jeu R 1

pif, s

.,.. s .s 1 -l [> j ) jQ -L 1 : ';l .z 3 4 y A/ .I .g. li l_ / ?I,- h IT l i l1 i ), 1 .r i... tid (M ~45) . fles (ECMS) ,1 ..i s t Figure 5 Plant Response to Feedwater Co'ntroller Failure ' (BOC8+8767 mwd /MT [BOC8+7953 mwd /ST) to EOC8)- Page 14 - 'g. .u ~ ~ 4

t 2mem i Pilgrim Rev.1 Reload ' 7 - i ? LPFl! C(Pii) 1p> r. x.(at rtur l' mv i-ser a mv tv ic seu roc. 1 t JA .l l... : t \\ ~a N N - N y - r.. en. .s. tid tucwsi- -l tid cucxu t

aygte.s,-ef g 7,4g,,,

iv .. t g iy35Ed'0* ,P * 'MU - ,f' =. y 1 ' i! 2 i 7 ..n ..g t b, y - h'. A 1 .. =. u.c cu:xs) vid cuexu p e f ~i

i l

l' Figure 6 Plant Response to MSIV Closure, Flux Scram Page 15 il

3 ki) u Pilgrim

(($ '

l Reload 7 1 0 - Appendix A increased Core Flow Throughout Cycle 8 3,1 The analyses performed for Cycle 8 included increased core flow throughout the cycle. There are no concerns regarding reactor internals pressure drop of flow induced vibration as- { ,l ' - discussed in the increased core flow analysis document for the EO 76 (NEDO 30242). The flow biased instrumentation for the rod block monitor should be signal clipped for a setpoint of 107%, since flow rates higher than rated would otherwise result in a ACPR higher tha q reported for the rod withdrawal error, i s S I i - i 4 .( f .t k i a .I Page 16 '- g (Final) I, i

l v 10CFR50.90 .p. 1 8057tW EDfSCW ~' Pdgnm Nuclear Power Station. Rocky Hdf Road Plymouth, Massachusetts o2360 l u Ralph G. Bird senior Vice President - Nsclear BECo 90- 136 ' U.S. Nuclear Regulatory Commission Document Control Desk Hashington, DC 20555 3 License DPR Docket 50-293 PROPOSED TECHNICAL SPECIFICATION CHANGE MINIMUM CRillCAL POWER RATIO Boston Edison Company proposes the. attached revisions-to Appendix A of Operating License DPR-35 for the Pilgrim. Nuclear Power Station in accordance-. with 10CFR50.90. The proposed revision'to,thesTechnical. Specifications upgrades the Safety Limit Minimum Critical Power Ratio snd revises-the. Operating Limit Minimum Critical; Power Ratio.- These proposed revisions will extend the use of spectral shift to maximize' fuel utilization during the present fuel. cycle. This change is.not,needed if our. proposed change of August 21, 1990'is approved first. . -) t $[f

t

, AA R. - ird RAH /njm/4820 Commonwealth of Massachusetts) County of Plymouth ) ] Then personally appeared before me,. George H.! Davis, who oeing duly' sworn,'did state that he is Vice President - Nuclear Administration of Boston Edison Company and that he is duly authorized to execute <and file the. submittal contained herein in the name and-on behalf of Boston Edison Company and that the statements in said submittal'are.true to the best of his. knowledge and belief. My commission expires: OM C/9W c2$4 M dv DATE ' - ]' NOTARY.P0flC Attachments: A. Description of Proposed Changes B. Replacement. Technical Specification Pages C. Marked-up Technical Specification Pages D.. Supplemental _ Reload Licensing Submitted for' Pilgrim Nuclear Power Station' reload 7 Cycle 8, 23A4800.RW, 1 1 signed original and 37 copies .cc: _See next page

BOSTON EDISON COMPANY. U.S. Nuclear' Regulatory Commission. Page 2' cc:1 Mr. R. Eaton, Project Manager Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Mail Stop: 1401-U. S. Nuclear Regulatory Commission; 1 White Flint North 11555 Rockville Pike Rockville, MD 20852 U. S. Nuclear Regulatory Commission Region 1 j 475 Allendale Road j a King of Prussia PA 19406 Senior NRC Resident Inspector' Pilgrim Nuclear Power Station Mr. Robert M. Hallisey, Director Radiation Control Program i Massachusetts Department of Public Health l .150 Tremont Street, 2nd Floor Boston, MA 02111 f , I

r Attachment A to BECo 90-136 l Qescrittian of Proposed Changt i Proposed Chances -Boston Edison Company proposes to upgrade the Safety Limit' Minimum Critical Power Ratio (MCPR) and revise the Operating Limit Minimum Critical Power Ratio. The revision to the Safety Limit Minimum Critical Power Ratio reflects i improved fuel designs in the core and the General Electric Coiling Water Reactor Generic Reload Fuel Application, NEDE-240ll-P. The revision to the Operating Limit Minimum Critical Power Ratio is being proposed to be consistent with the upgraded Safety Limit MCPR and to reflect spectral shift operation. Spectral shift operation results in' top peaked flux distribution patterns toward the end of the fuel cycle. A top peaked _ flux distribution pattern increases the change in the Critical Power Ratio (ACPR)- associated with the analyzed operational transients. The proposed revision maintains conservative operating limits. The upgraded Safety Limit MCPR and the revised Operating Limit MCPR ensure the plant will be operated safely and will not pose an undue risk to the health and safety of the public. These revisions will allow the. plant to be operated more efficiently. Basis for Chance The NRC approved the upgraded Safety Limit MCPR in its safety evaluation for Amendment 14 to NEDE-24011-P-A, " General Electric Standard Application for t Reactor Fuel", dated December 27, 1987. Incorporation of the upgraded Safety Limit MCPR is apprcpriate for BWR's with 0-lattice fuel assemblies provided: 1) the fuel has a beginning of life R-factor greater than or equal to 1.04 and consists of fuel types P8x8R, BP8x8R, GE8xBE or GE8x8EB; 2) the fuel is at least 2.80 weight percent U-235 bundle average enrichment;-and 3) the-lower i enrichment bundles in the core have operated for at least 2 cycles. 1 Thc fuel assemblies presently in the core of the Pilgrim Nuclear Power Station satisfy the conditions evaluated in NEDE-24011-P-A. These conditions will also I l be satisfied for future reloads. The Supplemental Reload Licensing Submittal for Pilgrim Nuclear Power Station, Reload 7 Cycle 8 submitted by BEco-letter 87.081 dated May:22,1987 has been revised to reflect operation using a spectral shift fuel management strategy. The revised analysis also reflects an assumption that the Turbine Bypass Valves do not open in the analysis of the Feedwater Controller Failure with maximum demand. This assumption was'made-because precise measurements of the opening time are not available. The assumption is conservative and it results in a more severe transient analysis and.a more conservative.0perating Limit MCPR. This limit is incorporated in the proposed Technical Specification. The proposed Operating Limit MCPR~will ensure the MCPR does not decrease below the proposed Safety Limit MCPR-at any time during any abnormalloperating transient- -l as defined in the FSAR. l

~. -. - _ - q petermina' tion of No Sionificant Hazards Considerations The code of Federal Regulations (10CFR50.91) requires licensees requesting an amendment to provide an analysis, using the-standards'in 10CFR50.92,.that determines whether a significant hazard consideration exists. The.following-i analysis is provided in accordance with 10CFR50.91 and 10CFR50.92-for the arcposed amendment to Pilgrim's Minimum Critical Power Ratic. 1. Upgraded Safety Limit MCPR-A. The proposed change does not involve a-significant. increase in the probability or conseq'Jences of an accident preViously evaluated because the proposed change to the Safety Limit MCPR does not change plant equipment, operating procedures, or governing design criteria;'Jsed to protect the plant against the initiation of any analyzed accident'or used to mitigate the consequences of any analyzed accident. B. The proposed change does not create the possibility of a new or different kind of accident from any previously analyzed because the proposed change does not change plant equipment, operating procedures,-or governing design-criteria and the change to the Safety Limit MCPR provides.the same level of protection as the existing Safety Limit MCPR against fuel cladding failure: during an abnormal operational transient. The proposed Safety Limit MCPR-therefore provides equal' assurance against a release of. radioactive i material in excess of 10 CFR 20 limits during abnormal operational transients and a new event sequence leading to an accident.is.not created, C. The following design requirement ensures an adequate safety margin-is maintained: Abnormal operational transients caused'by a single operator error or equipment malfunction'shall be limited such that, considering_ ' uncertainties in manufacturing and monitoring the core operating i state, more than 99.9% of the fuel rods would_be expected to avoid boiling transition. The propo;ed change does not: involve a significant reduction in th'e margin of safety because this design requirement,_which governs fuel cladding j integrity and maintains the defense-in-depth. philosophy, has not changed. 2. Revised Operating Limited MCPR-A. The proposed change-does.not. involve a significant increase _in the. l probability'or consequences of an accident prev.iously evaluated _because the proposed change to the,0perating Limit MCPR~does not change plant equipment, operating procedures, or governing design _ criteria used to l. protect the plant'against!the initiation of any analyzed accident'or used l to mitigate the consequences of any analyzed accident. l i _--_m,

t B._ The proposedLchange does not'createt the possibility of. a' new or different . kind oftaccident-from any. accident'previously evaluated because the j proposed: change does not change plant equipment, operating procedures ~,1or l governing design-criteria and the c :anges to Operating. Limit HCPRs provide the same-level of assurance that-the Safety Limit MCPR will not be exceeded ~ during an abnormal operational transient.thereby assuring'a release.of i radioactive material in excess of 10 CFR 20 limits will: not occur during i abnormal operational-transients and a new event' sequence' leading to~an- ] accident has not been created. C. The proposed change doeslnot involve a significant reduction in the! margin-1 ] of safety because the conservative Operating Limit MCPR ensures the most limiting. transient will not violate the Saftiy Limit MCPR. Reauested Schedule-j The requested change is not n'eeded until the present fueliburnup reaches'7200: MHD/T, about January 31; 1991. Additionally,> approval of;this request will.not: be needed if our proposed change of Augustc21. 1990 is approved first. l t i f l l .q l, h ~ I v

4} l Attachment ~B'to BECo 90-136 1 List'of Effective Pages l n . 'l Revised Pages s 6 1 1 13 i 205B-2 l 4 1 -f k s 1 4 5 ) l l:

+ 1.1 SAFETY LIMIT-2.1-LIMITING" SAFETY SYSTEM SETTING j 2.1 FUEL CLADDING INTEGRITY - l'.1 FUEL CLADDING INTEGRITY Anolicability: Aeolicabilitv Applies to the interrelated Applies.to trip settings of the ]1 -variables associated with fuel instruments-and devices which are j thermal behavior. provided to prevent the reactor 1 system safety limits from being-Obiective: exceeded. To establish limits below which Obiective: the' integrity.'of the fuel clad-- ding is preserved. To' define the level of Lthe process variables at which automatic, Soecification: protective ~ action-is initiated to e" prevent the fuel cladding = A. Reactor Pressure >800 osia and integrity safety limits from being: -Core Flow >10% of Rated- ' exceeded. The existenco of a minimum Soecification:. critical-power ratio (MCPR) less: ( [ than 1.04 shall constitute A ~. Neutron Flux Scram -violation of the fuel. cladding-integrity safety limit.' A-MCPR l of 1.04-is hereinafter referred ~ The limiting safety system-trip to as the Safety Limit:MCPR. settings shall be as specified = .below. B. Core Thermal Power Limit (Reactor-Pressure 1800 psia and/or Core

1. Neutron Flux Trio Settinas Flow 110%)

a. APRM Flux-Scram Trio When the reactor. pressure is- - Settina (Run Mode) 1800 psia or core flow is less

I than or equal to 10% of rated,-

When the Mode Sw.tch is in i the steady state core thermal Lthe-RUNLposition, the APRM power shall not exceed 25% of flux? scram trip setting design thermal, power. shall be: 9 C. Power Transient S'I.58H i 62% 2 loon. j ~ I The safety limit shall be assumed Where: to be exceeded when scram is. known to have been accomplished .S - Setting in.' percent of' by.a'means other.than-the' expected scram signal unless rated thermal' power (1998 MHt) analyses demonstrate that.the fuel cladding integrity safety H.- Percent of. drive' flow. limits defined in. Specifications -to produce a rated. core 3; 1.1A and 1.lB were not exceeded . flow of 69 M lb/hr. l during the actual transient.- i I Revision 6-Amendment No. 72 j --w-p +- "p - yew wi. g i

t j 1 BASES: The statistical analysis used:to determine the MCPR safety limit is based'on a model of the BHR core which simulates the. process computer 3 function. The reactor core selected for these analyses was a large:764 assembly,!251. inch reload core. Results from,the-large reload core. analysis apply for all operating reactors for all.-reload cycles.; including equilibrium cycles. Random Monte Carlo ~ selections of all operating parameters based on the uncertainty ranges of manufacturing: tolerances. uncertainties in measurement of core operating' parameters,t calculational uncertainties. W statistical < uncertainty associated.with the critical power correlations are imposed upon the analytical-representation of the~ core and.the resulting.bundletcritical~ power- -l ratios. -Details of this statistical: analysis are presented in Reference? ^ 2. B. Core Thermal Power Limit (Reactor Pressure' < 800 osia or Core' Flow - < 10% of Rated.),- The use.of the GEXL correlation-is :not valid ^for 'the. critical: power calculations'at pressures below 800 psig:or core flows less than 10% of rated.- Therefore, the fuel cladding integrity.safetyLlimit'is!.. l established by other means,. This is done by= establishing'allimiting- } condition of core' thermal power operation with the following basis. Since-the pressure drop.in the. bypass region is essentially all~ elevation head which is 4.56 psi the core pressure. drop at low power and-all flows will-41 ways be greater-than 4.56 psi.:. Analyses; show that:with i a flow of 28x103 lbs/hr. bundle' flow, bundle pressure drop istnearly' 1 independent of bundle. power.and has a value of 3.5 psi. ~Thus, the bundle flow with a" 4.56-psi driving-head will be greater thanj 28x1h3 lbs/hr irrespective of. total core flow.and(independent of bundle power for the: range of bundle powers of. concern. Full: scale ATLAS test. data, d taken at pressures from 14.7 psia to 800 psia. indicate?. hat.the fuel l assembly critical power at thisJflow.is-approximately 3.35 MHt.:'Hith: i the design peaking factors theE3.35 MHt bundle power correspondsLto a: core' thermal power of more than 50%. :Therefore'a core. thermal-power 1 limit of 25% for r(actor; pressures 1 below 800 psia, or-core flow less than 10% is conservative. u l^ 1 Amendment No. 42 11 2 - .c e w e s v v

4" d BASES: C.' Power Transient-i Plant safety analyses have~shown that theiscrams-caused by exceeding: any safety setting:will assure that the Safety Limit of_ Specification. 1.1A or 1.1B will. not be exceeded. Scram times are checked'. _. 4 periodically to assure-the insertion times are: adequate.=tThe' thermal, j L power transient:resulting when a scram is: accomplished other than by! the. expected. scram signal..(e.g.. scram'from neutron flux:following; ,J closures of.the main turbine stop valves) does not necessarily cause, 1 fuel - damage.- However,.for this' specification a Safety-Limit violation 1 will-be assumed when a scram is only accomplished by means:of'a backup feature'of the plant' design. The concept of not approaching alSafetyg-Limit provided scram' signals are operable is supported by the extensive plant safety analysis.- ~ ~ The. computer provided with Pilgrim Unit 1-.has a sequence annunciation ~ program which will7 indicate the sequence ~in which events!such!as scram, APRM trip' initiation, pressure-scram initiation, etc;, occur. This program also. indicates when the scram setpoint is cleared.-JThis will. provide.information on how long a scram' condition exists and thus o provide some measure of the energy added during'a transi_ent.-- D.' Reactor Water level-(Shutdown Condition) } During-periods when-the reactor is' shutdown,' consideration must'also be.- given to water level requirements due ~:to_ the effect of decay heat. = If" j reactor: water level should! drop below'the top ofs.the active-fuel during; this-time, the ability to coolcthe' core is reduced. This reduction in, core cooling' capability could-lead ~to' elevated cladding. temperatures and clad-perforation': The core can;be cooled sufficiently should the. water level be reduced:to two-thirds the core height... Establishment of' i the safety limit atel2' inches above.the top of the fuel provides adequate margin. This level will:be: continuously monitored. t References

1. General Electric-Thermal-Analysis ' Basis-(GETAB): Data, Correlation and

) Design Application, General Electric Co. BHR' Systems Departmenta November 1973 (NEDO-10958).

2. General Electric Boiling Hater Reactor Generic Reload' FuelqApp'lication, NEDE-240ll-A.

Revision Amendment No. 42 13 a A-s. L,,

i t TABLE-3.11-1:.. OPERATING LIMIT MCPR VALUES 4 A. MCPR Operating. Limit from Beginning of_ Cycle (BOC) to BOC + 7,953 MHD/ST. P8x8R/BP8x8R For all values of x 1.48 B. MCPROperatingLimitfromB0C+7.953HWDISTtoEndofCycle, P8x8R/BP8x8R t l'.41 t'_1 0 0.0 < t 1 0.li 1.42: -i 0.1 < t 1 0.2. 1.43-0.2 < t 1 0.3'- 1.44 l 0.'3 < t 1 0.4 1.45 0.4 < t 1 0.5. 1.46-0.5 < t 1 0.6' l. 4 7 -- 0.6 < t 1 0.7 '1.48: 0.7 < 2:1 0.- B 1.'49 if 0.8 < ' t 1 0.9 - 1.50 O.9 < t i 1.0 1.51- -l \\ e l 'l i I -l i Revision ] i Amendment No. 708 205B-2 i t

r c-t l i ATTACHMENT C to BECo 90-136 o Marked Un Technical SDecifications; 6 -i l i i I A F q -l 1 1 i 4 e

(

'I . i 'l

~ 4 9 _a 1.1-SAFETY LIMIT-2.1 LIMITING SAFETY _ SYSTEM SETTING 'l j 2.1 FUEL CLADDING ik'TEGRlTY-1.1 FUEL CLADDING INTEGRITY Applicability: _ -l Applicability: Applies to the interrelated _ Applies to trip settings of the: variables associated with fuel instruments and devices which are thermal behav'ior, provided to prevent the reactor- + system safety limits from being. exceeded. Objective: Objective: To definei he level of the pro ; ~ To establish limits below which-t the integrity of the-fuel clad-p cess variables at which automatic l protective action;is' initiated to' ding is preserved. prevent the' fuel cladding;inte - grity safety limits from being,' ' exceeded. i , Specification: Specification: -A. Reactor Pressure >800 psia?and

A.

Neutron Flux Scram: l Core Flow >10% of Rated q n 1 ,The existence of-a minimumi -i-The. limiting safety system trip! critical' power ratio (MCPR) settings. shall~be as: specified i below: ( j g less than shall consti-tute. viola, tion of,the fuel F 1. Neutron Flux Trip Settings-y cladding integrity etyJ limit. A MCPR of is'here- ) g s, -a.- APRM Flux Scram Trip-ine.fter referred:to as the t 4 Satety Limit MCPR. ,t Settina (Run Mode) l i B. Core Thermal Power Limit (Reae. 1 When the Mode?Swit'h'is'~ c tor Pressure 5800 psia and/or- 'I in the RUN position,' ) Core Flow 110%) [ the APRM-flux scram. 1 'll trip. setting shall be: When the reactor pressure is [ 5800 psia or core ilow is less 'S i.58W + 62%-2~1oop-y than or' equal to'10% of rated,- 1 the steady state core thermal i. Where: a power shall not ' exceed 25% of L j design thermal power. i S': Setting.'r perce:D of rated thermal C. Power Transient t poweri(1998 MWt) ~ W =LPercent ofidrive ,Thesafety[. limit-shallbeas a sumed to be exceeded when scram . flow'to produce a- ] y is-known to have been accomplished, rated core flow of1 i by a means other than the expected 69 M lb/hr. scram signal'unless analyses _ demon-1 strate that the fuel cladding 2 j . integrity safety limits: defined in... 1 ~ Specifications 1.lA and 1.lB were not exceeded during;the actual q q transient. s h h@ e 2 l s o

. = ll /MM i "A '" n ; r% 1..J. yvi i. G..iei.1esseel Je1 . e.-sne.ii.1.ilse- / lined n uth 5-1,Mhin:: 3, 2: r.'- ' :1x:: c' "' n:: j px;n:n; 1hted 2 tile b:, Lefs;-ee 0, d G..1stive--- e;; uily p.;; di n d W ;1es. ;L. L 71.I.e "-1 d b l. vi / 4 - A.I....ie 3r---tetries baA es.4 bla, Min z:. 3. Ci J i: 1 f e;n; dietetbett;n t'_ :-n; ' ;; " e - "_..e * * * * = ' 4 = 1 "..1-. .,,a .. m ...u... m. .y .~. ri v a. -distethet6ee;-- 2=. na uhn.=; ;M h;h i;; ;f -h !x1 ;; h. Th bnh fe m::t 4 th; i: St =n ;nr:tn; n; ;i;n -u 500 2031g rI th ta;1; in ;h.;.un,ehty la O -;;;;. 0 -eeereiesiee-60-3 tven-da-NEpe-1M58 W. n ;;rx dhnit_;'n u in.d n ; ;ir4-**l-ML 2 '17;;;;i:^t'd i: Rd ? ; t t ' --- .ee nitteet-ly-ebeoes :: gehn ; ;hn.2 p.-_. sh-di_.in i.t; -t h gr u n :: :-d u--ef n;"1he u ;h hisMG r-n 1---.b.

n. -

wee:: ft: trim t h-6-p&l;;in "ni n- % - n n: teem 'J:.it i fi--f.-- eey-hel--eyale n 1d n; L-se-eevert.e O. 41.616Livu.... i. Ibe

.lgi;.-

B. ,Q,,,m ner.tal power th.it (p.eaeter Pressure < 800 reir er Core T1ov ( 1 L. ef Fated) De use of the GD:1. correlation is not valid for the critic.a1 power calculations at pressures below 800 pais er core flows less than 10% of rated. herefore,gthe fuel cladding integrity saf ery id ' t is established by other maana. h is is done by establishing a limiting condition of core thermal power operation with the following basis. Since the pressura drop in the bypass region is escantially all elevation head which is 4.56 psi the core pressure drop at low power and all news vill always be graager than 4.56 pai. Analyses show that vith a flow of 28x.10 lbs/hr bundle flow, bundle pressure drop is nearly independant of bundle povar and driving head vill be graatar than 18x10ge flow with a 4,56 pst has a value of 3.5 psi. n us, the bund lbs/hr irrespectiva of total core flow and independent of bundle power for the range of buudia povars of concern. Full scale ATLAS test data takan at pressures fran 14.7 psia to 800 pata indicate that the fuel assee.bly critical power at thi.s flow is approximately 3.35 Wt. With the design paaking f actors,the 3.35 Wt bundle power cor-responds to a core thermal power of more than 5C! narafore a cort **.rmal power W't of 25% f or reactor pressures ban 800 i psta, or core flev less than 10: is conservative. \\ l ~~ b I Amendment No. 1

} Insert'h The statistical analysis used to determine the MCPR safety limit is based on a model of the SWR core which simulates the process computer function. The teattor core selected for these analyses was a large 764 tssembly, 251 inch reload core. Results from the large reload core analysis apply for all operating r.ictors for all reload cycles, including equilibrium cycles. Random Monte Carlo selections of all operating parameters based on the uncertainty rar.ges of manufacturing tolerances, uncertainties in measuremerit of core operating parameters, calculational uncertainties, and statistical uncertainty associated with the critical power correlations are imposed upon the analytical representation of the core and the resulting bundle critical power ratios. Details of this statistical analysis are presented in Reference 2. I l s i e l l C

i .. 1.:.1..:- .:.~ c_ ) .9 I l l \\ \\ C. Power Transtent i Plar.t saf ety analyses have shown that the scrams caused by ex-ceeding any saf ety setting viu assure that the saf ety Limit of Specification 1.1A or 1.17 viu not be extended. Scram timas l are checked periodically to assure the insertion times are I De tharmal power translant resulting when a scran sdequate. is accomplished othat than by the expected scraa signal (e.g., i scran from neutron flux fonoving closures of the main turbine stop valves) does not necessarily causa fuel damagt. However, f or this specification a Saf sty Limit violation viu be assumed when a scrat is only accomplished by maans cf a backup fasture of the plant design. na concept of not approaching a safety Linit provided scram signals are operable is supported by the extensive plant safety analysia. De computer provided with Pilgria Unit i has a sequence-annunciation program which vin indicate the sequence in which events such as scra=, AFF.M trip faitiation, pressure scram. initiation, etc. occur. 21s program also indicstas when the scras serpoint is cleared. 21s viu provide inf ornation on hov long a scran cendition arists and thus provide soma measure of the anargy added during a transient. D. Reseter Vater level (Shutdeve Conditioni During periods when the reactor is shutdown, consideration mist also be given to vatar level taquirements due to the aff ect If rasetor vatar leval should drop below the of decay heat. ) top of the active fual during this tima, the ability to cool. the core is raduced. n is reduction in enre cooling capability could laad to elevated cladding temperaturas and clad parforation. D e core can be cooled sufficiantly should the vatar level be reduced to t.ro-thirds the core height. Establishment of the saf ety if nit at 12 inches above the top of the fuel providas adequate j sargin. This level win be continuously monitorad. ) Refarances General Electric narsal Analysis Basis (CETAB): Data, 1. Corralatios and Design Application, General Electric Co. Novembar 1973 (NDN10958). 3VR Systems Department, 97 Precete -Oespter ?*rformanea-saluastem '-trey,- General- =:1;;tri: C ;rty 5'N Syet:-- 0:;nrrtr-Jume 494-- r

s. _...,.

General Electric boiling Water Reactor.Genaric Reload ,/ d Tual Application, NCI-24011-P. h 13 e Iu

I i l TABLE 3.11-1 OPERATING LIMIT HCPR VALUES A. MCPR Operating Limit from Beginning of Cycle (BOC) to BOC + MHD/ST. P8x8R/BP8x8R 7,953 D/.48 For all values of t / B. MCPR Operating Limit from BOC + MHD/ST to End of Cycle, j / 1 P.SxBR/BP8xBR ) f/ / /. 11 0 kW / e 0.0 <t1 0.1 he /, Y2 0.1 <11 0.2 h/.M3 0.2 <t 1 0.3 h/.MY t 0.3 <11 0.4 he /48 / h /eMb / 0.4 <t! 0.5 0.5 < 11 0.6 /Y?

0. 6 < t 1 0. 7

/eYO 0.7 <t1 0.8 -h43 /, Mf // s 3 ^ 0.8 <t1 0.9 he /. 60 f 0.9 <11 1.0 he /, f/ ,/ / i I Revisionh 205B-2 L W /b l

,..~.. ~.. .y, ,.~ E~ j. i '. Q& <t-s f (i, b.., +, s ..41 t r L(' ,w. i,= .f- .! r-'i s > t j- :: i y 1 F ; m, ' i g 3. ., ( . so j i. 4 j,- ?; s s,, V;4 * ,,r 4

. k, j =Vl A -%.. / p. 'q +

4a ,^ a i, 32q 3 <,. = e., il, y.d o b ,~<, -v . g;. ,,1 - ! l." W J , 5:, , -. r.t 5g;' ' d*a-r -..}.. 3piU -. +t ? Wu .I

We

! t n , s ' A. ' i r 4 4 1 W '.1' r { p. ':a-5 - 3 S q 1 43 T L p' / cf 3 i y 1; ( s f- ' ' i%,. 'J'y2 7 J., a 7j = A sum o ..o'. ',,, K:. i. x, y%y : h ' r;T '; ; f .g , 1 1 y 4 r . n w', 1,. M, 5jj.( a

t. 4.{ l A
'gf '

A 2 ' [ M, 'ki y. I - 5'I-e", s i %o... ."o a, s s, maq _m' - .,~.,,,g, t E I ,m{ /, / j I 2'. ' i) e_ 6 I l 7 l ,'t pq3e w< n.,; x_, a,,.. v ;. - w-

4> I4,.

v ,3 7.a. , J '.,,. p. w,. r 3r t 1 <. t 3 t - w .r- +y

a.. ' q 64

=a g _ 4 4 +,.. a.(i -s - s + 4 iATTAC10(ENT;D!toLBEco 90-136D 3 1 + WS' ..;5 3; .x Pw 1 + 1 o -r. i y 'y m-y. 'i <0 ,y s r,4 e J' s f' 3 ..,i' - d p, i a f s ig '1 7 r>V. 4 e s 4 J'e. t :' t k 1' b j b-7 ( I' '1,- 1 s~h' b. g ( 3? g ..f : a )- 4 g VI y;b= ..{ _.y-s + ,e v u S. a s i g ,n w ~,

~~L,,,.-,.~~
~~~.s v f 9...' + e y ~ 1 __ A v f T., + .' e fI g 1 T g g a 3 t -j- ,I 4 1._ I r .._l-.w- @ * > 1 .a o ). q- -s -:d. L 3 -y g ,g '] I - I f. { g., ,f. ,3-3 h 5~~
~~j E~~
~~6 t ..y e;~ mh' L f' i - . ) [: Li; 3 4 )$ 4 s _t h2 3. 1 L, s s, -. c.. # _ f +g +r ' i t 4 .s < .,g .-n 9 d r i t a] v,s~~
l'
~~[ -ry..j s 7' 4 i~~
~~4 a~~
~~'.s J.'p* - 4. ; ,>a 7 s r i e p,. '7~~
~~x I~~
~~)'t' L.ry? ;,, ;-, 5 n + ~ r 4 y[; '; b' g _;; g 5 ib 5 'g, , m ,t, +, ,s.~~
~~W4_,~~
~~S4 .a, ~j) t -i u n s ! [: t -. a3 ' i A.. - + f y7 b b f 4 . h g ;g q -y ( ) t ( L 4 } 3 6 n m h i f ni t~~
j%

o.
~~es + i U _ i _a; +y',- v. 5 1 kf 1 h{[ t .,1'._ 1 ,o lp (.; ? -l1 ~w f, r. + c ie t -- L 3 A~~
~~, n s~~
~~4t i b s s sg ,y, 4, [ I $ b.h[ vh k* y. ), J,, o, i i* '?< --Nr, J. 4 -1, ,.) , g. 32 h e -r 1 [ a 5 4*w m,a } t g P k " } y , -.. 8 g g -t t~~
~~g. r a~~
~~1 'qt= ' i r7' 4 " ) i a J C-I' .,t m w, ,e 1-3 e + 4 i 1 q s, ...m i ,*3 '.E .3. g 3 .',-1,- , d.*- 9.,,. f [4] f -. ( 9 F l\\ T i-g 9 ) ..l, 6 - N 6 i al } g J h A e Ih~~
~~u.. t f f~~
~~y.[ w - ? t' 4 a - v -t t ev < j +, ] '4 -(p s 1- > ) e, g h i 5 W k ,) .,,,w A l-t' g" ' I, V: B J '.5'^ s* !g'o ,) '. ~ .y - e5 4 1 e g '.e ( >a e j 'f'~~
h
~~'U' i' ( 6 3 I x -t: e y q l _f'. I-( j i %g1, >.8.f .-l. r s v. 4 1 c I q i?: "\\ Ph-~~
~~1 p~~
~~e s~~
~~5 t a~~
~~g . -li. -s t _ p p w , _' h, y'95 s I T ' w'A n. l Jf '*d.- ( r g p .] 4 .l -a. - y 3 4' ' 7 + ,r t~~
~~) 'i,,~~
5, 1 t ~ t ..} 7 3 i .e-i y y: .a { 5 Y', d. l-,, ) r o s r. i p f ,'t E g'. T I ,7-g 's4 9'< 1 s r =.1 tl y. 4 t u t ( .[ 5 l, y 3 1 1.. 1- .m. 49, gn .g " (j g Wt ' Q-a t 55 Vu s t... ,3r 1 g.- g.., h :., Q :.
~~f. : _,y, ji~~
?
~~6 s~~
~~i n... s_ = mo _ o, [ 4# c m k s-- 7 g w1 .c 'M as,- I ~ y f { -f. h' -- I[ -- y d t_ g, e[. N m' a,~~
~~a,..,.~~
~~ji g .. n , %g w,,,u t,ne,v,. .. ~ : n . # e-br4 5, e., w.s o w a,, g Ms .%,,ew w w w %w,w w,, om,w,%.,% ,,.w m., ,,,y~~
~~g. m m v., w &p wrw rwee c w~~
t I CE Nucker Energy l 1 23A4800 [ g lieilslon 1 5 Class I - September 1990 l 23 A4800, Rev.1 Supplemental Reload Licensing Submittal l ~' for Pilgrim Nuclear Power Stction i Reload 7 Cycle 8 I I i f '1 ^ / a, M Approved: \\'jkl Approved: / . J. S. ha ey, thager E.
~~ansen, Manager Fuel Lie sing Reloat: Nuclear Engineering~~
,k l g-l
e i Pilgrim max Reload 7 Rev i i Important Notice Regarding l Contents of This-Report Please read carefully This report was prepared by General Electric Company (GE) solely for Boston Edison Company (BECo) for BECo's use with the U. S. Nuclear Regulatory Commission (USNRC) for amending BECo's operating license of the Pilgrim Nuclear Power Station. The information contained in this report is believed by OE to be an accurate and true representation of the facts ~! known, obtained or provided to OE at the time this report was prepared. The only undertakings of OE respecting information in this document are contained in the contract between BECo and GE for fuel bundle fabrication and related services for Pilgrim Nuclear Power Station, and nothing contained in this document shall be construed as changing said contract. The use of this information except as defined by said contract, or for any purpose other 8 than that for which it is intended,is not authorized; and with respect to any such unauthorized use, neither OE nor any of the contributors to this document makes any representation or warranty (expressed or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infrinp privately owned rights; nor 8 do they assume any responsibility for liability or damage of any kind which may result from such use of such information. t l l Page 2 g
4 EIlgFInt 2M4MO keload 7 g,y i j i Acknowledgment i 1 The engineering and reload licensing analyses which form the technical basis of this Supple. l mental Reload Licensing Submittal, were performed by E. G. Thacker II of the Fuel Engineering 1 Section. The Supplemental Reload Licensing Submital was prepared by P. A. Lambert and i verified by G. O. Jones of Regulatory and Analysis Services, i r i t I r I i l l 1 i h 4 Page 3
i .4 i Pilgrim max } Reload 7 kev 1 -i 1. Plant. unique Items (1.0)* i Appendix A: Increased Core Flow Throughout the Cycle .l 2. Reload Fuel Bundles (1.0 and 2.0) =! .i Fuel Troc _ Cicle Imded Number j Irradiated P8DRB282 5 24-I P8DRB265H 6 60 { P8DRB282 6 112 BP8DRB282 7 32 P8DRB282 7 160-1 New BP8DRB300 8 19.2 I Total 580 l 3. Reference Core IAnding Pattern (3 2.1) mwd'/MT mwd /ST Nominal previous cycle core average exposure at end of cycle: 18,362 16,658 Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 18,362 16,658 -l g Assumed reload cycle core average exposure at end of cycle: 21,018 19,067 Core loading pattern: Figure 1 1 g
( ) refes. to area of discussion in General Electric Standard Application for Reactor Fuel, NEDE 24011 P A 9, September 1988; a letter "S" preceding the number refers to the U.S. Sup-plement, NEDE 24011 P A 9 US, September 1988.
I g Page4'
\\ 4 Pilgrim mnm l Reload 7 Fev i i 4. Cgculated Cere Effective Multiplication and control System Worth. No Volds, i 20 C (3.2.4.1 and 3.2.4.2) Beginning of Cycle. K,,,, Uncontrolled 1.116 t Fully controlled 0.962 Strongest control rod out 0.990 R. Maximum increase in cold core reactivity with exposure l Into cycle, AK 0.000 5. Standby Lignld Control System Shutdown Capability (3.2.4.3) f Boron Shutdown Margin (AK) (DDm) - (20'C. Xenon Free) 675 0.040 6. Reload Unique GETAB AOO Analysis initial Condition Parameters (S.2.3) i Exposure: BOC8 to BOC8+ 8767 mwd /MT (BOC8+7953 mwd /ST) Fuel Penkinc Factors Bundle Power Bundle Flow Initial + Design Lccal Radial Min] R Factor (MWt) (1.000 lb/hr) MCPR BP/P8x8R 1.20 1.62 1.40 1.051 5.454 _108.6 1.44 Exposure: BOC8 + 8767 mwd /MT (BOC8 + 7953 mwd /ST) to EOC8 Fuel Pokine Factors Bundle Power Bundle Flow Initial DnigD LOCD} Radial Mial R. Factor (MWt) (1.000 lb/hr) MCPR BP/P8x8R 1.20 1.59 1.40 1.051 5.355 109.2 .1/7 7. Selected Margin Impmvement Options (S.5.1) 1 Recirculation pump trip: No Rod withdrawallimiter: No Thermal power monitor: No 4 i Improved scram time: No ~ I Exposure dependent limits: Yes I Exposure points analyzed: 2 g Page5-l
~~s Pilgrim-zwar i~~
~~Reload 7 Rev 1 j I 8. Operating Flexibility Options (S.5.2) - j 4 Single loop operation: Yes~~
~~1. cad line limit:~~
Yes Extended load line limit: Yes Increased core flow: Yes Flow point analyzed: 107.5 % Feedwater temperature reduction: No ARTS Program: No j Maximum extended operating domain: No 9. Core. wide AOO Analysis Results (S.2.2)- i Methods used: GENESIS Nuclear with GEMINI ODYN and OEXL ACPR Flux O/A [ Event (% NBR) (% NBR) BP/P8xRR Figure Exposure: BOC8 to BOC8+8767 mwd /MT(BOC8+7953 mwd /ST) Load rejection without 651 125 0.32 2 bypass Feedwater controller 663 132 0.37-3 failure Exposure: BOC8+8767 mwd /MT(BOC8+7953 mwd /ST) to EOC8 Load rejection without 730 126 0.35 4 bypass Feedwater controller 718 134 0.40 5 failure I = t-hae 6
r i Pilgrim zwam I Rev 1 Rgload 7 i i
~~10. Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary l~~
~~c ric ounding Analysis Results Rod Block ACPR Readine BP/P8x8R 104 0.13 105 0.16 106 0.19 107 0.22 108 0.2P 109 0.32 110 0.36 Setpoint selected: 110 l~~
~~11. Cycle MCPR Values (4.3.1 and S.2.2)~~
Safety limit: 1.04 Non nressurization events: 1 Exposure range: BOC8 to EOC8 BP/P8x8R l Loss of feedwater heating 1.21* Fuelloading error 1.23* Rod withdrawal error 1.40 4 Pressurization events-Ootion A - Ontion B ' BP/P8x8R ' BP/PRx8R Exposure range: BOC8 to BOC8+8767 mwd /MT(BOC8+7953 mwd /ST) Load rejection without ' 1.42 '. bypass Feedwater controller 1,48 failure 'The minimum MCPR value required by the ECCS analysis is 1.24. " Option B is not available for this exposure, Page 7 g l l
f 4 j Pilgrim 2wsm Reland 7 Rev 1 i i
~~11. Cycle MCPR Values (4.3.1 and S.2.2) (continued) s Pressurization events:~~
~~Ontion A Ontion B BP/P8x8R BP/P8x8R 1 Exposure range: BOC8+8767 mwd /MT(BOC8+7953 mwd /ST) to EOC8 l Load rejection without 1.45 1.40 bypass i Feedwater controller 1.51 1.41 failure~~
~~12. Overpressurization Analysis Summary (S.3)~~
~~Pd Pv Event (psig) (psig) Plant Resnonse MSIV closure (flux scram) 1317 1333 Figure 6 4~~
~~13. Imading Error Results (S.2.2.3.7)~~
~~Variable water gap misoriented bundle analysis: Yes Event ACPR Misoriented fuel bundle 0.19 l~~
~~14. Control Rod Drop Analysis Results (S.2.2.3.1)'~~
Pilgrim Nuclear Power Station is a banked position withdrawal sequence plant; therefore, the control rod drop analysis is not required. NRC approval is documented in NEDE.24011.P.A.US, September 1988. 4 I I ( Page 8 4 l -~.
I t Pilgrim new i neload 7 m e, i i
15. Stability Analysis Results (S.2A) g Pilgrim Nt, clear Power Station is exempt from the current requirement to submit n' cycle specific stability analysis as documented in the letter, C. O. Thomas (NRC) to H. C.
j Pfeiferien (GE), Acceptance for Referencing of Licensing Topical Repon NEDE NO)1 Rn'. 6, Amendment 8, 'ThennalHyiraulic StaldlityAmendment to GESTAR 11,' April 24,1985. Pilgrim Nuclear Power Station recognizes the issuance of NRC Bulletin No. 88 07, l Supplement 1, Power Oscillations in Boiling Water Reactors (BWRs), and will comply with the recommendations contained therein.
~~16. 14ss of coolant Accident Results (S.2.5.2)~~
LOCA method used: SAFE /REFLOOD/ CHASTE See Loss of Coolant Accident Analysis Report for Pilgrim Nuclear Power Station, NEDO 21696, August 1977 (as amended). 1 v 'l 1 i f Page 9 i
Pilgrim - m a:o l ll M BsM M BE M BE MBEM M M M M M M BEM BEM M BsM M BfM BE
MMMMMBEMMMMBsMM

MBsBBBBBsMBsHMMBEMBE 1

BsBEMBEMMENEMMBsE

Bi M M BE88 M 88BEB8BEBsM Bs

BsMMBEMMMBEBsBsBsMBE
~~'::BsM M M M M B8BEMEEEsM M i~~
~~MBsBEMMMMMMEsssMM M M BE M M M BEBBBEEE E E @ E_E @,_D Q_0 0 0 @_0 E 0_E Q_E 8~~
d@@@nen@dsesFedscs 1 s +8 @YY0E0 2 E l l l l-l' l l I l-l l l l l 1 3 5 7 9111315171921232527293133353739414345474951 FUEL TYPE A = P8DRB282 D = P8DRB282 B = P8DRB265H E = BP8DRB282 C = P8DRB282 F = BP8DRB300 I Figure 1 Reference Core Imading Pattern Page 10 il
) Pilgrim: i nua neland 7 mev 1 ) i i 4 ) A* I$r$["g.t rw, pMh50 2 [t t
~~&{lF5')~~
~~emm m i L A f f,,,,: ' v(h. / N f k l \\ ~ m. i I. ed. r22 i .i :~~
q.
taw cucxt> tiw (scoes) hfh!fc. ,/T ! f fj" x m. i. y A ' ' '. N I"- w/ , wm-ii j 9 \\ 1 ia s. i. taw (ucees) tow tucees) I .f. Figure 2 Plant Response to Imad Rejection without Bypass (BOC8 to BOC8+8767 mwd /MT[BOC8+1953 mwd /ST)) l' Page 11 l4
Pilgrim m46x + Reload 7 an. 3 f k ses.. rtyi h afh hf' Nv h ra
~~.(si f.4!!!ts i[ ii-RfINI P tl~~
(
~~i..~~
~~r i l...; "." b I-e-4 I 3 5 .( ' t . s s,, L 9 tid (MCWCl) Tid (K 04%) f f f~~
~~2 3~~
a r. ~ = = t f E '( _3 2 or u.. Ild (MCDOS) tid (Etec5) 1' '] i i Figure 3 Plant Response to Feedwater Controller Failure ' i~ (BOC8 to BOC8+8767 mwd /MT [BOC8+7953 mwd /ST)) Page 12 .. j
Pilgrim m4sw Reload 7 kev 1-( ji, ! TOM "- @fe id""" 4 uh /- 6 - 4 yx 7 [ N x, 1 4 1/ :. .g . Ild (MCM4) tid (ECM$) fJ"5 / TAII'i I M M E"S"!!!" ifdf% fit!F M "< 5'") / - ~V = x a f) s -'}g w e s c }f .., r: I .. s. 8%. . tid (n.MS) 16d (ECMS) 1 8 Figum 4 Plant Response to Generator lead Rejection without Bypass (BOC8 + 8767 mwd /MT [BOC8 + 7953 mwd /ST) to E6C8) r.se o
Pilgrb avs* Reinnd 7 ket 1 8 1*fa'[%Mht la{s:n!{17J h M %'"') rw. n s1 + l5 l \\ -e l a.re g \\ L m. ii< cums) i < <a u.esi dWi2iF"" lNN, in / W .;., y 4 E kl n t_, e... e ,j 1 .i. I ile (MCDCS) ind (Eceet) f i Figure 5 Plant Response to Feedwater Contmlier Failure (BOC8 + 8767 mwd /MT (BOC8 + 7953 mwd /ST) to EOC8) Page 14 l-1 u-
8 Pilgrim zwam Rpload 7 Rev 1 .'hrN"* l h " h. f hk m. Y i,... Jf% m N f Nx t =.
~~3 r.~~
.s. .=. i. lid (MGM6) lag (MCet$) i ' [g g ? [fglCP b Ri) ,$m.,J., git.11v.j,iiM,T! t 4 i 4r M.. [2 (, r m,, J j' =. ..c .i p s -~ L Ip E g y \\ Mc. t ~s. s%. tid (uCxs> tis tuc esi 4 .i L Figure 6 Plant Response to MSW Closure, Flux Scram i Page15
5 i Pilgrim m. Reload 7 gn i Appendix A 3 Increased Core Flow Throughout Cycle 8 i The analyses performed for Cycle 8 included increased core flow throughout the cycle. t There are no concerns regarding reactor internals pressure drop of flow induced vibration as ' discussed in the increased core flow analysis document for the EOC 6 (NEDO 30242). 1 The flow biased instrumentation for the rod block monitor shoeld be signal clipped for a l setpoint of 107'7c, since flow rates higher than rated would otharwise result in a ACPR higher than - reported for the rod withdrawal error. k 4 I y g. Page 16 9, (Final) L ~
t 10CFR50.90 g sosMWannKW Ngnm Nuckat Fbwer station l Rco y HM Road Pignauth, t/assachssetts 02360. Ralph G. Bird senior Vice President - Nucleat BECo 90-136' U.S. Nuclear Regulatory Commission Document Control Desk Hashington, DC 20555 License DPR-35 Docket 50-293 PROPOSED TECHNICAL SPECIFICATION CHANGE HINIMUM CRITICAL POWER RATIO Boston Edison Company proposes the attached revisions to Appendix A'of Operating License DPR-35 for the Pilgrim Nuclear Power Station in accordance-with 10CFR50.90. The-proposed revision to the Technical Specifications upgrades the Safety Limit Minimum Critical. Power Ratio and revises the Operating Limit Minimum Critical. Power Ratio. These proposed revisions will extend the use of spectral shift to maximize fuel utilization during the present fuel cycle. This change is not needed if our proposed change of August 21, 1990 is approved first. Ti pe.-- 4~,d 'B rd . A A. RAH /njm/4820 Commonwealth of Hassachusetts) County of Plymouth ) Then personally appeared before me, George W. Davis, who being duly sworn, did. state that.he is Vice President - Nuclear Administration of Boston Edison Company and that he is duly authorized to execute and file the submittal contained herein in the name and on behalf of Boston Edison Company.and that the statements in said submittal are true to the best of his; knowledge and belief. My commission expires:- O M C / 9 98~ clI4 M i DATE '- }' NOTARYP0pIC Attachments: A' Description of Proposed Changes B.-Replacement Technical Specification Pages-C. Marked-up Technical Specification Pages D. Supplemental Reload Licensing Submitted for Pilgrim Nuclear . Power Station reload-7 Cycle 8, 23A4800 Rev.-1 'l signed original and 37 copies' L ,cc: See next page J l.
1 BOSTON EDISON COMPANY f U.S. Nuclear Regulatory Commission Page 2 cc: Mr. R. Eaton, Project Manager Division of Reactor. Projects - I/II. [i Office of Nuclear Reactor Regulation Mail Stop:- 1401 U. S. Nuclear Regulatory Commission 1 White Flint North .i 11555 Rockville Pike Rockville, MD 20852 l U. S. Nuclear Regulatory Commission Region I 475~Allendale Road King of Prussia, PA 19406 l Senior NRC Resident Inspector Pilgrim Nuclear Power Station 7 Mr. Robert M. Hallisey, Director Radiation Control Program Massachusetts Department of Public Health 150 Tremont Street, 2nd floor Boston, MA '02111 i t i
i Attachment A to BECo 90-136 Descriction of ProDosed ChanQt i Pronosed Channes l Boston Edison Company proposes to upgrade the Safety Limit Minimum Critical Power Ratio (MCPR) and revise the Operating Limit Minimum Critical Power Ratio. The revision to the Safety Limit Minimum Critical Power Ratio reflects improved fuel designs in the core and the General Electric Boiling Mater l Reactor Generic Reload Fuel Application, NEDE-24011-P. The revision to the Operating Limit Minimum Critical Power Ratio is being proposed to be consistent with the upgraded Safety Limit MCPR and to reflect spectral shift operation. Spectral shift operation results in top peaked flux distribution patterns toward the end of the fuel. cycle. A top peaked flux distribution pattern increases the change in the Critical Power Ratio (ACPR) associated with the analyzed operational transients. The proposed revision maintains conservative operating limits. The upgraded Safety Limit MCPR and the revised Operating Limit MCPR ensure the plant will be operated safely and will not pose an undue risk to the'hehlth and safety of the public. These revisions will allow the plant to be operated more efficiently. Basis for Chanat The NRC approved the upgraded Safety Limit MCPR in its safety evaluation for Amendment 14 to NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel", dated December 27, 1987. Incorporation of the upgraded Safety Limit MCPR is appropriate for BHR's with D-lattice fuel assemblies provided:
1) the fuel has a beginning of life R-factor greater than or equal to 1.04 and consists of fuel types P8x8R, BP8x8R, GE8x8E or GE8x8EB; 2) the fuel.is at least 2.80 weight percent U-235 bundle average enrichment; and.3) the lower enrichment bundles in the core have operated for at least,2 cycles.
The fuel assemblies presently in the core of the Pilgrim Nuclear Power Station. satisfy the conditions evaluated in NEDE-240ll-P-A. These conditions will also be satisfied for future reloads. The Supplemental Reload Licensing Submittal for Pilgrim Nuclear Power Station, Reload 7 Cycle 8 submitted by BECo~ letter 87.081 dated May 22,<1987.has been i revised to reflect operation using a spectral shift fuel management strategy. The revised analysis also reflects an assumption that the Turbine Bypass Valves do not open in the analysis of the feedwater Controller Failure with maximum demand. This assumption was made because precise measurements of the opening time are not available. The assumption is conservative and it results in a more severe transient analysis and a more conservative Operating Limit MCPR. l This limit.is incorporated in the proposed Technical Specification. The proposed Operating Limit MCPR will ensure the MCPR does not decrease below the proposed Safety Limit MCPR at any. time during any abnormal operating transient i as defined in the FSAR. i
1 J Determination of No Sionificant Hazards Considerations j 4 The code of federal Regulations (10CFR50.91) requires licensees requesting an amendment to provide an analysis, using the standards in 10CFR50.92, that determines whether a significant hazard consideration exists. The following analysis is provided in accordance with 10CFR50.91 and 10CFR50.92 for the proposed amendment to Pilgrim's Minimum Critical Power Ratio, 1. Upgraded Safety Limit MCPR A. The proposed change does not involve a significant increase:in the i probability or consequences of an accident previously evaluated because the proposed change to the Safety Limit MCPR does not change plant equipment, operating procedures, or governing design criteria used to protect the plant against the initiation of any analyzed accident or used to mitigate i the consequences of any analyzed accident. B. The proposed change does not' create the possibility of a new or different kind of accident from any previously analyzed because the proposed change does not change plant equipment, operating procedures, or governing design criteria and the change to the Safety Limit MCPR provides the same level of protection as the existing Safety Limit MCPR against fuel cladding failure during an abnormal operational transient. The proposed Safety Limit MCPR therefore provides equal assurance against a release of radioactive material in excess of 10 CFR 20 limits during abnormal operational transients and a new event sequence leading to an accident is not created. C. The following design requirement ensures an adequate safety margin is maintained: Abnormal operational transients caused by a single operator error or equipment malfunction shall be limited such that, considering-uncertainties in manufacturing and monitoring the core operating state, more than 99.9't. of the fuel rods would be expected to' avoid boiling transition. l The proposed change does not involve a significant reduction in the margin of safety because this design requirement, which governs fuel cladding i integrity and maintains the defense-in-depth philosophy, has not changed. 2. Revised Operating Limited MCPR ~ A. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the proposed change to the Operating Limit MCPR does not change plant equipment, operating procedures, or governing design criteria used-to protect the plant against the initiation of any analyzed accident or used to mitigate the consequences of any analyzed accident. 1
( i B. Th, proposed change does not create the possibility of a new'or different l kind of accident from any accident previously evaluated because the proposed change does not change plant equipment, operating procedures, or governing design criteria and the changes to Operating Limit MCPRs provide the same level of assurance that the Safety Limit MCPR will not be exceeded during an abnormal operational transient thereby assuring.a release of radioactive material in excess of 10 CFR 20 limits will not occur during abnormal operational. transients and a new event sequence leading to an-accident has not been created. C. The proposed change does not involve a si!)nificant reduction in the margin of safety because the conservative Operatung Limit MCPR ensures the most limiting transient will not violate the Safety Limit MCPR. Reauested Schedule The requested change is not needed until the present fuel burnup reaches 7200-KN0/T, about January 31, 1991. Additionally, approval of this request will not be.ieeded if our proposed change of August 21, 1990-is approved first. h t F s i
. nw. +4 r M.M@m'T I,( p e ; # a,' b - y t sa nd.1 m m "m). n.. f.,... q, },h Ep't w
~~+ n ~.~~
~~~ m,u.,;, t~~
~~m. :) }, J = ' -* I i, pr lig iikI~~
)[ p'f9- . c ~4> W W ' A@wC ? ZW. T@M G?M Wj t%a m'M-4... g x x, n'lf i, ;, y n* *= -.f' m[r y ?' ~ . je m' v u' ' ' i j,_ ; }p.'. i w M K -i cg p ey - s. Q%.v yy)( s 'm w w'~, 3' h .2: y,, ogp j h '4 um P i- ?mX CN tw o%=sM . ~ a-t- ( 1' Wy%wp,s f. 3 c t C,o #!x Mmai @, TRM-OMM$x.$v A Mb@WW"" %, Z,M}$m$; t. M" @ Ns M - W igf"W JW 'S. +. ;b g..L.%o'I
~
~~= ,, I c,~~
~~m. n., m 3wm c_~~
~~s - w m,w m~~
~~o., t" mn. rm.~~
~~h 31s~~
3
~~'D m.we, s 7{ c'.. ' 'T _m4.?.n-n 2 g y q %< a5.. ..m m-h.2; _v. - 1e:- y a, y m mn ,w<~~
~~y<1 s,~~
m, a u p.w- "1 1 2 y~ ':'.s f,, R. -u c y A,n if-pif { C'-, A- ' " .Jg l 4-e,.c..*
J
~~' u d' ' d.; m m'; { .-l'; . N ' ' Q;c' q t -/ < m e_>y,tMih j )y{NL -.s s 3 -,h ct -hr -s ._s~~
~~f. t~~
-^
f = N' Q' = $. e.' LD.f a n' k{g% j @' kQ M% [ - ': [ %, + ]' =f'< b' ' h j]p . g.p g~g rL. wL j g!;&,Q[' '$% 7.y, . i v . p.+. ,'s fe!;;r f,b ?< MO. m#*_-'u r$f:Q& n, j fy _ -_ - r:l$-~V9 ~G t m_ ' i a, ' rf C # M 6, %> W Y_a} V - T W }, v , n y o r, .w - -e< 4 ! Attachment 8 to'tECo 90-u W s. .j m.. -. # s. m 71 x, m y v. s '- ~ .,m ,g. O y F4-e.4.+. 7, . n n_ LJq d L. i $n,;e-" s W x, 4 i t I. : ry 1< 7" 4 x..v., 7 e s U' st of fffective.: P. ages?w ~,g m+n.f *. ".A, . t+ W $o . a 4 o e + s. , tw a.b m: 1, m n w4.. - - 3., >z 4 yt.) -4
~~- -4 1 yv7~~
4.. g%s

%s,,n.
i. - A ; sp< r pty -, e p. = t 4 2 c 2 1 b YE N. s +a.s a _ g..f, id +# 1p,b. ;- e u -. + m m, 'I-t'.I Yh! + AI... -m 1 r l 'i 5 4 s e <<$.I we.g-a qy c, y ag s. 4 2. y,u-.. A' y_ j c .. y;. d" [w$>& eg,. A,$s $ - w = , @,, m qe Mit C ..g-_,k n p. w, m w 'I --q. q:M o a v n Mg J dDE '; v M i @ 'W h. i ' -y, s mA -, , v: n i" g4g(p g,, p p-. gg,, g;.%m -- " .n S5 A-A 84 % i i 4 i is.- + s ph s - j,4-?- O .J 4 pt n, s,v,. < i m .g @q w. gg _ g' f ;'.g } g- : h re 7 3 ;4, w- -g 4 ,N i.} i_ g in 4g ,4 ,. g hy -a, ; 4 '( r 3g'. g, 9
~~a -~~
t-
~~;. y j;y -~~
~~t apg ;j, 3 c g,m%w.g y,.r,,og 3_,y ;y i ,. - g~w - a.. o 4 1 m w b _/y[f~~
's
'I" . ' '. ~ g a c y y 9, f ' { * .y a y,. k g 5 - ) y) Y f + '_'-'{_, y 4.h m. l \\ ,a s... 4, -., i t h I 4 2% y z [ j,,., {_ b I m. s 4., s %u. w p w'. 7 y-4 3 - %m.y y. i - m N I ,{( ] $1 ,k4 a f;-t 1 g,"- =, r.g: i- !v; j v ' - m n v x.w' ww a.gewa. n + {s!'
~~W J~~
~~- t -. r' z. / ',t o i / Of \\ @,ENp.. 4 -.sh, -,1 e-y" C' 4 A,s O & T, L M Q .,m -te r y*J + g Y, y j.. ..:_-.., -st,- -) b). t - 1 r en " = a - ht ,s k -.~~
~~T'%~~
.o 4 mf L. n L' ' '.. <b{. 7 y u-t.. ' 'v n9 mwd. - ~. e w ,c. c- .r: -> 4 ma,,# 'i- .y ,t + i au .+ -4 e,r '1 m, s, s I /- -- %%;ut;. wa y.; O. +4 [ . - - A, em r * ).~ I' [ h, N.f i b-f'i#1 1 3 a= a n @. }; n w'n'f _ , q- _. na U. s v~
f,

a. l v g3 -
7g y -N g.; ,et t. j f p'.. n@; s_ u a " i ,t 4 n- ,.4.. L-43, y y c3 _e_ e to", e r uw . c,,p o .-.' /,4.3i.. %} g. ' =. n n y1p4 g wg l{-f I y. # a-,3_, , e_, c _.c.J j t i J, l t j a p' s I y r t', c' i_7 Vj i. g .o ; i F. " y
~~4. [ C _~~
~~4/_ 1 2 ,x 1 js ( M .r',, g _,, E.Y.. av" t ,I- - m,3, 3, ', r, r ..'J. ..t [- {. -[ ( --3 l.( ; 1 I J E L. ".qg .g,pr [' .m g a z%. - ^ 7 ;y.y~~
~~m. -_Y -.~~
~~in s 4.P " I i ,1 r jh s. . - "[( x ,(= a,- -, f.* m o b}1 1 q.' 3 E'd 3- - - -'[ ; r 7 ..'{',i 3 ~*Ir J ' ,4 1 ,l5 ). }- t [ g "' i 3' y- (~~
~~% M2~~
~~-Q'[ [h=. j6'p w :, ; s ,4~~
~~g y.~~
~~4 s L,. F a_ p....gqq 3 g h .-m ....:, c. . y, s, n I P~~
)
~~j 4 j e' i i y 4 Y : f, T 5 ,l' 'f r - "c, F_ --O p i, ',n~~
~~ipc~~
~~. p. ' c j 'i l}L.,, ' %. I -, l I ~ b ~# 4 z. g, - v /t: ' v^ . t i 4:, L k n n, o ,1 ,. ' n,. +d{ ('ge , :1; ' i \\ .= - ,Lk g i-[ og m c,. "p ~ t 4, f~~
~~r. 3 - ',~~
~~t Ni> .y 9 I 4 T Y ,{.~~
~~f. -~~
~~i 4 3 4 ' v.. e t f~~
~~"_3> 13 i~~
~~1'. 5 h". d...# r + 9 _'0269 m&, e L 9 t n + o-i 1 9 Qg a. p ..% y. - 7 W 'y ; }; r 9;, r , + c i=. 3, yx > b _f' g g' 4 s I <' ). I T g 3 1 . f. s',hl t t .;, j i.,; 3 -]~~
~~s e, o~~
~~.y +, o s. 4 t eg; 3 t . p: ';1 - 4 J ; 4 1 t 4.. t j k. g 1 i+,, h,*.,,.b - 3 'I - l' 2 ,,,= i2 =' _, -. '{,.. h,n " ( ' ? [. g v i.~~
~~l. y.~~
~~= > ,) i u y. j '( &'y.. s Yg f. ' s i-' s e, t r. 1 . wa A c(s. 4 .Av q, s ( + s 1~~
~~h' E~~
~~I qt + f. 's xa. \\~ -..4, e{g j % JL ,.[' p' !"' uW, 're T&4aFm ',, y,.~~
~~p v.~~
~~a 2 < :M h. Vi,. '", er t 4 , a;-~~
~~t. 46_~~
~~g t -...t .,c ,a, y v g. 1 ' k { k 1 [} 'h o ~ g 4 3 6 S 4 g* fi A ?- r 1 r, c;. 'R ii - m' ~ + i 'o"g,.:p sc'y udM s s , g,~~
~~.L ?~~
4
q-
~~+ 4 e - i,, ,) st e + zg , p,: ?.; . ;$E.. r 4 r m.. wm s; y' yy k, .t 4' A. V 1 ...?~~
~~t. l_ U~~
ka g Y L %,_m,'-( ( 'O ,]i +, h \\ -.
~~y s m.~~
~~r 1: U', y 'y g' / a 5 i<. s s ,. ny sp g g = g 3 / g4 k' l-:'. n 4 vu - u_ ? v:b t1." i, V x(,- w j f-g g,, Y vW ' S} 6 i, s \\~~
~~a. '~~
~~o F t t. . ;g. 1 ( 4 -- p O -. ' g ' T. A 4 ,p., y,- 7 9 J'Oi s bi 5 7 -dE ',i p i4 5 t M 3t (9 (~~
~~
p p.t.m M.hy. j.... < s 6' _MM u - .&. If,E f.p o.D,i '.%Y :. Rt c u... O-3 y y& ~ + m s f( r -y&e? Y g w, \\ g Y s ,n %f ~, t c,n, n, ym.., n _.N., 'I ) t . x. v.u 2., a-, g, e t n g ' g.r.m?, s',.\\ n:< -Q f nj p N k' }%' ?.% 'y.3,,v ; h y' p,- f'h i hR N Y.. 4 T* i - '_, s ', ;, __ _ ) l f. ' $f' h'k' - Esg m-iy ?* + r 1 fd'= i h'? o 4 4m3 ,, ' ^ k,, n.,.,.., 1 4, ' Nh *[b $,,._. 4 i;h,Em f.
~~.g 3~~
~~)s u; flT& ' lj rN', Y ( I~~
~~< - 7J Sj~~
.t ',t h,A f?-.m" ?g g { l / ][igQp $m$$@.,f'YfQg g%yl?g N: ff, r-, [ wfW],l(( ]< dm l _ ? '.yq _ Q?h, i,f. > k f- ,,,Q g m s-4 $ %.f; ggg g 1 a m Q# f yg g j$.s i i
1.1 SAFETY LIMIT 2.1 LIMITING SAFETY S(STEM SETTING 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Aeolicability: Aenlicability: Applies to the interrelated Applies to trip settings of the variables associated with fuel instruments and devices which are thermal behavior, provided to prevent the reactor. system safety limits from being Obiective: exceeded. To establish limits below which Obie'ctive: the. integrity'of the fuel citd-ding is preserved. Td define the level of the process = variables at which automatic Snecification:. protective action is initiated to prevent the fuel cladding l A. Reactor Pressure >B00 osia and integrity safety limits from being Core flow >10% of Ratd ' exceeded. The existence of a minimum Soecification: critical power ratio (MCPR)'less 1 l than 1.04 shall constitute A. Neutron Flux Scram violation of the fuel cladding integrity safety limit. A MCPR l of 1.04 is hereinafter referred The limiting safety system trip to as the Safety Limit MCPR. settings shall be as specified below:. B. Core Thermal Power I M (Reactor ? Pressure 1800 psia hnd/or Core
~~1. Neutron Flux Trio Settinas Flow 110%)~~
~ a. APRM Flux Scram Trio When the reactor pressure is Settina (Run Mode) 1 800 psia or core flow is less l than.or equal to 10% of rated, When the Mode Switch is in the steady state core thermal the RUN position, the APRM power shall not exceed 25% of flux scram, trip setting design thermal power. shall be: C. Power Transient S 1 58H + 62% 2 1000 The safety limit shall be assumed Where: to be excee.ed when scram is known te havo been accomplished 'S = Setting in percent of by a means 'Jiher than the rated thermal power expected scram signal unless (1998 MHt) analyses demonstrate that the fuel cladding integrity safety H p Percent of drive flow limits defined in Specifications. i to produce a rated core 1.1A and 1.1B were not exceeded flow of 69 M lb/hr. during the actual transient. 4 Revision E LAmendment No. 72 e. a
7 l' BASES: s - The statistica1 analysis used to' determine the MCPR safetyslimit' is based on a model of the BHR core which simulates the process computer = function. LThe reactor core selected.for these analyses was a:large 764 -assembly, 25111nch reload core._ Results from the large reload core-analysis apply 4 for all operating' reactors for all reload' cycles,-- including: equilibrium cycles;. Random Monte Carlo selections,of all Loperating parameters based on the:uhtertainty ranges'of manufacturing. tolerances, uncertainties in measurement of core operating parameters,, calculational uncertainties, and-statistical uncertainty associated with-the critical power correlations are imposed u)on the' analytical representation of the core and the resulting aundle cr_itical power ratios. Details-of-this statistical; analysis are presented (in Reference; 2. ~ B. Core Thermal' Power Limit (Reactor Pressure < B00 osia or Core Flow' <-10% of Ratedb The Use of'the GEXL correlation ~is 'not valid for the critical power. calculations at pressures below 800 psig'or core flows less than 10%'of rated. Therefore, the fuel cladding integrity safety limit ist established by other means.1 ThisLis'done 1,y establishing a limiting condition-of core thermal power operation with the'following basis. t Since the' pressure drop in the bypass' region-is essentially all elevation head which-is 4.56 psi the core pressure drop at low power and 1l all flows will: lways be greater than 4.56. psi. Analyses show that with 9 a flow of 28x10 lbs/hr bundle-flow, bundle pressure drop is;nearly independent of bundle power and bas a value of 3.5 psi.. Thus, ~ the r bundle flow with. a 4.56 psi--driving ';iead will' be_ greater than 28x103 lbs/hr irrespective of total' core flow and independent of bundle power for the range of bundle powers of concern.. Full scale ATLAS' test data 1 taken at pressures from-14.7 psia to 800 psia indicate that the fuel [ assembly critical power at this flow'is approximately 3.35 MHt.- With + the design peaking factors the 3.35.MHt bundle power corresponds to a _j core thermal power of more than 50%. Therefore a core ~ thermal power-limit of 25% for reactor pressures below 800 psia, Lor core flow less than 10% is conservative, l l a i i Amendment.No. 42 12 1 1 l w _ __u
BASLS:. C. Power Transient .j Plant safety analyses have-shown-that the scrams caused by exceeding any safety setting will assure that the. Safety Limit-of, Specification? 1,1A;or 1.1B will not be exceeded. Scram times are' checked-periodically to assure-the insertion times are adequate.- The thermali power transient:resulting when a-scram isfaccomplished other than by' .the expected scram signal (e.g.' scram from neutron flux following closures of the main turbine stop valves) does not necessarily cause fuel damage. However, for.this, specification.a safety Limit _ violation will"be assumed when a scram is only accomplished by means of a backup-feature of the plant design. The-concept.of not. approaching a Safety 1 Limit provided scram.: signals are operable is supported by the extensive plant safety analysis. q j The computer provided with Pilgrim Unit 1 has a sequence annunciation I program which_will indicate the sequence in which events'such as scram,L 'I APRM trip initiation, pressure-scram initiation, etc., occur. This-program also indicates when the scram setpoint is cleared. This will provide information on-how longLa: scram condition exists and thus-provide some measure of the energy added during a transient. D. Reactor Water Level (Shutdown Condition) j During periods *. hen the reactor is shutdown, consideration must also-be given to water icvel requirements due_to,the effect of decay heat.- If l reactor water level should drop below the, top of'the active fuel during-this time. the ability to cool the coretis reduced. This reduction in: core cooling capabi,ity could lead to elevated cladding' temperatures-l s and clad perforation. The core can: be' cooled sufficiently-should.the water level be. reduced'to two-thirds the core height; ; Establishment.of the safety limit at 12 inches-above the top of the fuel provides ~ adequate margin. This level will be continuously monitored.- References
~~1. General Electric Thermal Analysis Basis.(GETAB)i' Data, Correlation and.~~
~~Design Application, General Electric Co.sBHR Systems Department, November 1973 (NEDO-10958).~~
~~2. General Electric Boiling Hater' Reactor Generic; Reload Fuel Application, NEDE-24011-A.~~
~~~ i \\ C Li Revision . Amendmont No. 42 .13 .F~~
l
~~N 1~~
~~TABLE'3.11-1~~
~ OPERATING'. LIMIT MCPR VALUES j t' A. MCPR.0peratir,g Limit from Beginning of Cycle.(BOC);to BOC + 7,953 MHD/ST. P8xBR/BP8x8R For all values of't 1.48 B. MCPR Operating Limit from BOC + 7,953 MHD/ST to End of. Cycle. -P8x8R/BP8x8R t t 10 ' 1 '. 41 ' O.0 < t 1 0.1 1.42 0.1 < t 1 0.2 .1.43 0.2 < t 1 0.3 ,1,44-0.3 < t 1 0.4 1.45 O.4 < t 1 0.5 1,46 0.5 < t 1 0.6 1.47 l 0.6 < t 1 0.7 1.48: 0.7 < t 1 0.8 '1.49, f 0.8 < t 1 0.9 1.50 0.') < t i 1.0 1.51 l 1 l-. t ' Revision Amendment No. 108 -2058-2 l! l 1
-- 1 \\ '1 i t ATTACHMENT C.to BECo 90- 136 - Marked Un Technical Specifications i - i [ \\ ,l [ f i ~ ! k I 'l 1 . i ..l 4 k i i-e y
4 P l.I' SATETY LIMIT 2.1 ~ LIMITING SAFETY SYSTEM SETTING j2.1 FUEL CLADDING' INTEGRITY-1.1 ~ FUEL CLADDING INTEGRITY Applicability: j Applicability:: S , Applies to. trip settings of the Applies to'the interrelated variables associated with fuel' instruments.and devices which are-1 thermal-behavior. provided to' prevent,the= reactor-l system' safety limits from being 1 ~ ] }' ' exceeded. -) _ Objective: Objective: To establish I'imits below which. To' define the level.of the: pro-f the integrity,of the fuel clad - cess variables at_which automatic-protective action -is initiated to - 1 ding is preserved. i prevent the fuel cladding inte-grity safetyjlimits from being' l 1 exceeded.- ] Specification:
~~Specification:~~
~~( A. Reactor Pressure >800 psia-and - 'A.~~
~~Neutron Flux Seram Core Flow >10% of Rated The limiting-safety system trip-The existence of a minimum i~~
critical power ratio'-(MCPR). settings shall.be as specified j, g less than shall consti, .below: l tute viola, tion of.the-fuel cladding integrity fety. 1. Neutron Flux Trip' Settings. 1imit. A MCPR of 1s-here-gg inafter reierred to as-the L a.- -APRM Flux Scram' Trip i Settina (Run Mode). Safety Limit MCPR. B. Core Thermal Power Limit (Reac-J When the Mode Switch is tor Pressure 1800 psia and/or I in;the RUN position,- Core Flow $10'.) .i ,the.APRM.Ilux' scram. -i' trip setting,shall be: When the reactor pressure is i-5 800 psia or core flow is less ~ .S.S.58W + 62% 2 loop q/ q t 3 than or equal to 10% of. rated, the steady state core thermal' iWhere: L power shall not. exceed 25% of j designLthermal powet. i S 'e. Setting in percent of rated thermal C. Power Transient h l power _(1998 MWt). R i W= Percent of drive. 1 The. safety limit shall be as-sumed to'be exceeded'when scram flow to produce a' ] is known to have been'acecmplished? rated ~ core. flow-of 1 <69 M lb/hr. by a means other than the' expected scram: signal'unless analyses demon-strate. that the fuel. cladding s a = integrity safety limits defined in., ? Specifications 1.1A and 1.lB were not exceeded during the. actual ~ transient. l. 6 Amendment.~No./ u
.j i + ' // // Th; re v1..J u,,wi ;., e..L..ie.1 col del . e. _. .s i ' cites-- / 1:aud ;; Obh !-1, *:!::=n 3, 2: = '=1
1

1-r! &c =::
~~i p n ;; t;;e-st ::d-i; O tte b l, Eifer.;.ee 0,.od.e.~~
~~1. dv.~~
~~j is "i wre. b 1_.ud b 1A vi / i / 7~~
~~,;;;;.t17 pean di n -itatie; ehv.~~
~~E.fese.e O. Teklee bM e i b2", Ref;;;;n 3, in th.. E-fe n s distelbeti;;; '10~~
~~ir ut 11 9 eteti"" *"'. =.^d*'~~
.,a..- 1s-1rset __w ..wm .u. .u u.; m.. w n. w. ...uva -41stelbwe4::: d;.. ete-eeseen-see: de h;;ir.;.i;;; ;f 2: ".:1.7 1:. -75: b;:1: fe: de r:::: '-ti:: i: de n:: p;;=::n: n: gi;= l 1; 0:00103L0(2) ;;g g; 3;;;; ;;7 7,g;,,3,; 77z> = ;7 ;; g, = -ee reeletten-to--given i; C10?5 sui. "'t; p=:: di: trit.;i;; is t;,;;d ;; ; ;gic;l 754 ;;;--ily :::: 6 -ti;h $2 ::d p; : = Wee ed i n arily ch =;; te produce e ek s.d pe e: din rih. tin.' #.;E F --th; g:nn:t - 'n ;f.;;-- ' ides et 0; highen peee: Lev.le. The
~~di ::ibe:i= in P11;;-1; "=h;r !=:: i::t1;; ";i: 1 dri;;~~
j f=_1 j;1

=.14 an be es e,...re Ce di...;Lw;1vu.. 4 iu m - __,_
_m m B. Core Ther=al Pever Lini: (Reseter Pressure < BOO psic or Core Tiov l <10 ef Rated) The use of the GEIL correlation is not valid for the critical power calculations at pressures belov 800 psig or core flows Therefore,4 e fuel cladding integrity-th less than 10% of rated. saf ety li=1t is established by other means. This is done by: establishing a limiting condition of core thermal power operation ' J with the following basis. Since the pressure drop in the bypass region is essentially all elevation head which is 4.56 psi the core pressure drop st lov i power and all flovs v111 always be graager than 4.56 pai. Analyses show that with a flow of 28x10 1M !hr bundle flow, bundle pressure drop h nearly independe:: of bundle power and drivingheadvillbegreaterthan28x10geflowwitha4.56ps1 has a value of 3.5 psi. Thus, the bund . lbs/hr irrespective of total core flow and independent of bundle.pover for the range of bundle powers of concern. Full scale ATLAS test' data takan at pressures from 14.7 psia to 800 psia indicate that' the fuel assembly critical power at this flow is approximately 3.35.MWt. With the design peaking factors.the 3.35 MWt bundle power cor-responds to a core ther=al power of more than 50%. Therefore a i core thermal power lir.it of 15% f or reactor pressures -belov 800 ' psia, or core flov less than 10% is conservative. i \\ i I 4 Anendment No. 12 r I + e...,..
l i Insert A* l The statistical analysis used to determine the MCPR safety. limit is based on a model of the BWR core which simulates the process computer function. The reactor core selected _ for these analyses was a large 764 assembly, 251 inch reload core. Results. from the large reload core analysis apply for all operating reactors for all reload cycles, including equilibrium cycles. Random Monte Carlo selections of all operating parameters based on-the uncertainty ranges of manufacturing tolerances, uncertainties in measurement of core operating parameters, calculational uncertainties, and statistical uncertainty associated with the critical power correlations are imposed upon' I the analytical representation of the core and the resulting bundle critical power ratios. Details of this statistical analysis are presented in Reference 2. 8 A 1
.;.r.:. = ~ -r m L o. 'f. g-I j C. Power Transient Plant saf ety analyses have shown that the scrams caused: by ex. cl ceeding any saf ety setting viu assure that the Safety Limit of Specification 1.1A or 1.15 vill not be exceeded. ~ Scram timas J are checked periodically to assure the insertion times are - De thernal. power transtant resulting vben a scram adequata.- is accomplished 'other. than by the-expected. scram: signal' (e.g., l scram from neutron flux fonoving closures of' tha main-turbine stop valves) does not necessarily cause fual damass. Bovaver, f or this specification a Saf ety. Limit' violation vin be assumed vben a scrar. is only accomplished by means cf a-backup fascure of the plant design. ' ne concept of not' approaching a Safety Li=1t provided scram signals are operable.is supported _by, the .i e.xtensive plant aaf ety analysis. t ne computer previded with Pilgr:.a Unit 1 has a sequence - ~ annunciation program which viu indicate. the ' sequence in.which, events such as scram, APRM trip initiation, pressure scram-initiation, etc. occur, his program also indicates when the-scra= setpoint is cleared.. n is viu provide information on' hov long a scra= condition exists and thus provida some naasure-of.the energy added during a transient. l D.' Reactor Water Level (Shutdows Condition)- i During periods when the reactor 1s shutdown, consideration.nnist - L also be given to vatar level requirements due to: the affect: .[ If raacter voter level should ~ drop below the j of decay heat. top of the active fuel during this time, the ability to cool the core is reduced. his reduction _ in core cooling capability .i could lead to elevated cladding temperaturas and' clad perforation. The cora can be cooled sufficiently should the water level be; reduced to tao-thirds the core beight. Establishment of the'saf ety limit at 12' inches above the top of the ' fuel provides' adequate margin, nis-level win:be continuously nonitored. - Re f erences_ General Electric Thermal Analysis Basis (CETAB): Data,; 1,. Corralation and Design Application, General Electric' Co.' November-1973 (NEDO-10958). 3'IA Systems Department, ) --m 6 6 ::s 0 ;;;.:t:r ?crf:-- :: i 9 -ti n Mr" 2 y, C rreals
~~12:tri: 0 n;rry 3'a Syctrr- 0:p r r rt, Jer, 1976~~
': 0 0 000'^). ,. + / General' Electric Boil!.g Water Reactor. Generic Reload '/ d yuel Application, NEDI-2l.On-P. 13. r ...~....... r I., .-;~ 4 L-
.t TABLE 3.11-1.. OPERATING' LIMIT MCPR VAltlES- ] A. HCPR Operating. Limit from Beginning of Cycle-(BOC)'to BOC + MHD/ST.:- 'P8x8R/BP8x8R-3953 D/.46 For.all values of t- ./ - B. MCPR Operating Limit from BOC + MHD/ST to End of: Cycle.- / / t P8x8R/BP8x8R" 10 /$ll h/,Mf, 0.0'<t1 0.1 h/.dl 0.1 <t1 0.2 h/.N-0.2 <t1 0.3 hb48 / 0.3 < t1 0.4 0.4 <11 0.5 /I d [ 0.5 < 11 0.6 . /. M b / 0.6 < t1 0.7 /N i 0.7 <t1 0.8 4-6 /,3f / 3 s-0.8 <T1 0.9 he+ / 60 ' j ' S'/ ,/ 0.9 < t1 1.0 hM- /,- l I h l lI '.Revisionh 2058 > W fle.
MMMn.,&wwy%;M:W,w 1 J ? ; : a %WWML%3+%g?Mb n~ e ......-+ng-- W %,WM4 M M + 2 M,~ W &n % g W h Y hh lh h M @ % E W W. U@%;#%, M$yAQMy@@ @$nglMWW wy@MW $Y?% h* h Qh - +pl' yp M Q:%Q/e[ w"%((ed & : w( e(s - i] h. m% ' t wm v.. ww + w , ~ 4 - s[ c - -I'% {. . ew -{w wjj-p hh - e G' 114.j [ h (. Md - g li A T 3. J.
~~u y~~
) .,.m -sv ye m.m. _ %;.. M %w v1x y< . v a m mm r >-e 74%~^~ wW,M.. %.W W ;,;ATTACHMENi.sD4to"BEcol,90-130 1 CW @r 1 N, 3r 4 Y !Suppleme.ntal Ael_oadtLicensA.. & M_ 9g av m ing,: Submit tod d or; Pilg, rim 1Nuclears Power y Wwr@ 5. m ym Mms e > og y e "..s. _ a m . m i g pwu M. - W:- J, n .t. y m mmmc m aM' M N Y UN 4 Station 3eloa'd3 Cycle:8M23A4800$RevMID 4 ' mkWr QMN t 32 y' , k_; Q, s. w u _ e,y%, 4.,, - o u p o y,p.,Q w& w9 ~. -,t, w, a o g ,1 .% ma -- w ' + - dog e a Q,,.,;, c g m x .g. Tg,; ~ .3 -s g e s
~~n., i a~~
~~..g y &;'jM Ni _y .r y n. 5 l y_' $y m,. Qi a ~ y Q. ..., ',}l ( fm^ ' ;.-mu ,. x w,'~~
7

1 7__
~~hi 5 ~% Eb N$,.,s' m ~;s s w w w A~kDf -_ n.,w. -h,,.' ~ sp4* w t - .w_ v I'm ' ' ,~~
~~e 5 's.5~~
~~'^l, ll" -l e,,- + c: 4 ,.u..~~
~~T.;~~
~~" $ " ]' ' ? ? W}M } [ D'f, > l9gf 'M' m i .,%' Q:i -4 c3 i-r r '. :A s~~
~~jd s~~
~~A -fwYSr ij J , $' s t h ' ', 'hu ' ,t~~
~~$mc thM,- -@aJ.,E. e hu SM'.- '~~
~~b~ 1 PU=% \\ . " y,. m['n ! W i n . ~ , s A%f%- v c. , f.. rW _4 4 - mng sa2 q t 4 x -j.+'s a 1 N ,,s'- .t.- [ - em - e- -Q dgl y t, i - cm 4y3 -;ydl,-j. g a; ufN m< m -Q: v g, _qQqy r - w~~
~~w-w i~~
s ~ . w.t. m, 4 [ . i, c.b T AL ?> ";#M,C.y X. . ;n 1.--g4iv) R,D,C' @h ;G.R.y .G. s .wg cf." 2.r'3.. 4 v-4',.,p._ y,,J,.'. .t .-M' ;.. '..l?e; _.. f E. y.c,t ,(d- _.h[ll+,'y[. - '-.f F ,dE .3",- ,I ..,..}qy ,'gry
(
3 - p j e g t - a n, B; {g. , 4 [; y 'Q < > j'. # Q;[ s + yy% j ,LTc') j y e g;a%pqggyu,W-kJ4R /p: Q QQ_ ^ u murr } Q-e_ n' &s 4QL_ ; t ^~z w_. ^ r> y v w @,i-;;Q r, hp "'l i l7 ^ r J y, g, y { t t + 9 e 4 yh ca D:: q& q r 5 ,p %y + +.. 1.p,
~~4., r~~
~~z M po a s --~~
~~y, xp = & e -" ea . g! ' ' ?.dt b+g ' ii 3 "-@h '% nh n= i ,v k_ c_ i t +* Ni ?.? ) W' M.. 3 g41 A ^ 4 e A Tc r * %. M '&'r~~
~~14..-~~
~~cr +, &,s 4 4.~.%* k. = < -: v.E-,#.b ' 'a .( i a~~
~~y.,.~~
y a. v fi,t ? 4 x1 i-- +,/", ,i l ^ [~.,./c 3tC4- 4 F ' s q.q'4 40' t 'a'w: - p.w 9{41'.s. s i"1-t? s n._. A. 4 L g.'t 1 . z' cg ga : r3 S c @, -- i % { - i. 2 .i' t -d:, y 97 g; x, '![ p &.g q. g s y ;;pk f %yrm y n;-p; V-- t q' g v '. -U Rpsm, o m s an y
~~i. f i fe, %~~
Eby U 1 6 ._s M f ";. >, w r y; ed ,,V q r c
~~p. Q im +~~
~~-J ; ? ,c w y,, q e u v {;, _ ~_ e $p. La !p s,fh Nf j y e r y 5 n ,, o! Vj l. r' j f j;, oW ' ' ) c,, kf_ ff' 4f? b a nn 1 = # W. sp.. a a. s~~
w
~~? M.5 p w..d t/ r NRW' i ~ ;g@ N* " 4W. Z R i s(e , (,,,. + t =, 4~~
~~3., g E M7 = : -~~
~~t -77 y'l~~
~~.7t:Q.3sst t~~
~~Sp ?; f Wp% <iq MMi n e -s +4 Q pir ' 'q' " ' - y . gh F " 9 ', y s t y ,3, g 4. e3 @u .. M . p-n.g.v b. s q,.~~
~~4 w:~~
~~2. 5.r. 1 ~ y J. w~~
~~m. w<~~
~~m 1 q.: t~~
~~. s3 bpe eg,, 't.~~
4 + s s y s .n.. i .Y# e i' cm > .m. <,%i s t g.A b x--4,.' q'- ('t ,,1 ,,,3 p j (1. s;j 'y ' + m1 -(_ 4 j a n 4- - b p .y-- 0 b.> ,_y y W* t , ' ' ' ' '.',j ! I r f -6 ~3:. ~, .Q',- s s r4 s g u ' ..y 'o '.,.[a E 4 -'l Nj j_ o f e w' 4 .w-" g. fa .i b? s !';n 4,S. N, mW? GM M .w. ab .. i >MO r0. W< a ry gf h D_'> k 4
~~' y.g g~~
~~s 4 . %>y,i,xqa, w n+< 4 _, s y; uA, a W d, r3aw w. m.m n .,a. m; A a e' +- )w;.4 ' 4 2'a#* e,4 'I'4 t Y 3 +'yP "4.'..: s i _4 p, i 1l. - 4 !g.- .:? j; kn b},' ;. 334 gw -f gig w' agi~~
Qf
~~_\\ g-i s-M. ,.r,.s ~.p-;d yr e1( t-, y) 3 g. 4 .p e-3 3 t g r,ya %j y ... t ; v w,u ,g, M,WK f, .m. m. b%... * . - 4R , %:q m%,l, Rf. w -2 .fp'. Y f.--bx 1 =',~~
~~m. 'm.~~
~~4 m },a2 k jh ! Y ' I'e-x b. p 4 y), 'N N 4 r' 4 .n g.w'y wp.' >4, :~ n, - y =,,. p,y._ 4 5 i..u4 s~~
~~gV1 d~~
~~i 4' k. f' g? f A k m, - x a _qv p -: ~ 7 9 -t t y y a.:, s. r .. i le t. - ng.- M 1, W 5 69 l t h P- ~ 2 J 'M. 21 i@y by9@@g.J $ 'i.p' (%',,.. p. s / + S.G vd>,I?? t~~
~~3 9~~
~~. pf, SfOK t "t~~
~~y.,O;;~~
~~.-t i s m ..~3 g q pg_ d; p y, 45$w q T f [s.~~
~~f he%u@ (y ',~~
~~g m s ' $ * # M% M, 7N _ f mm f ~h m.p m~ m. ~ ) ~ M M' N{- % '- p c ss wme e ~ d jr' y4 a 23' 7 ',+,,:_k f . i ,f<^; 5~~
"

4
~~%'l.Qfi @ M 'O..j% y iA7'd @ Qv 12" !. y> iW; QN;* 4 Q ' ]t + > Ip w~~
s/
~~'M ..N,,' /1~~
~~v. s~~
. A; ki 5, B i m w=~ s -3 m qu. x m* %??' '%UF'* % s.f. 5 vf l ~ ;M rd. p gI ' :?. a. i i \\ Nf C., A i, e, -; \\h V.' i og 4.', L.M y m. "34 ,.=(_. 5m y ,. ;g< - ich.A < 4 5 r> ?3 - .sppg m ,, m d d; :;f_ . t, %g,;;fn , pyQ k 8 p 4 1 4 g 4. ^' s
d

..
.;., w, L;m s. s -a, acq. r -h_-,_,. %y y a n%( i Y I d b. ,4-^ i 6, )t '.; E, d. I f m 1 p m. [... u e c Tr,m =* pg J 3. g y z p y ;, : G.v-fW ~ 4 Li ), ,. y,q.. s 30 e}-.. y 1 e, s, w' ; da,, - t W; T -., v_, <r i , ui as n y,, + .r me mc ^ip. a. r -@1.* I (y fp:6 n F %%U
~~n, ' <. p,. y%m %n,,~~
~~(, v &,' #$, ' ;: a .o. A j. e'm y m + 4 wk ,.,y;.." . L L nn L- ~. 3~~
~~nn t~~
~~A, h d % &e .g ^ Q, a f. F s y u k.g~~
~~s 1 3,,- ~,m w~~
~~u. s ge -m ..v,s m a n y. < 1 , y m y ~l ~ $i s l f } i h f h 4 + g. a f 1 .3 t:3 %'~~
~~v. p :,. 4 3:c..~~
( v v s ,?.. - N $' l' < !,' Y iN bW! L A '; L 'T ' 3(.. 'g),r 1,4 ,a =. g y' ,;g ? '. IL" 7' t r .o. s' i ,3 s x 'w en s f# M+ n n,3 l o,. i;.. ' "'.. < 9) n.f W4 $,f.. 1. s).L r .~q {MW~ jfj 'dy t s j (t; ' % Q QV.,, Q;f,fy V '_~t >=~.v. t > s s e, r ...s: .L. -:p.'
8
. dS r + -L i g es w - s u.a b 's, J, _ s v, ' 1 L W, e < dw 8,.' ..( 4+f,.. p, . =. ' 3$1 ',. t p$ s 5.. 3 y -- ?-3 i q.. 1 ( i t- . i i t .g f q s..; p x i + .s. w _ gg, '? t... ,,x 1 12 1 P. s e :t s q q^4h ,h, ' m. '~5u.I u; I ~@'.. 4,,. i( -,.'[' -.-[ f /,4 i h 7[ i [! i s' i an iu 3:4 w .... -i,,i. - w 't 1 1 _g y ':; i
+-
~~~', i (J M W-%N~~
~~7s$i-4 n n~~
'; F. g $)h ' a- ;m. )-[. w ) ] 'ij,,q ' [ [- 5 5 kqN! 7 mew: _--i. j . c 4 +-- > = up - ec w 3 '. s, s t 4 o huen,? 2 w m'. ; am a E-S 4 FM L-M! A w n q? A r w v <- J s m= gy' n H* - 3: er n 5 mM si-W T W.W,, - N ~ t-A r '.W 4,, h JA .r
y'
~~~ 2 gb fx : s.- h.? m~~
~~d. -~~
~~gj M k[' Ig, kh. b N::- h '., ?e -) 3, i f%g* '+) 2 1 1 m gr - UU n w w %a er y. 4: r t +mt M.. o m wp j tmA s' o nA m e >i as un...i s w-g b~~
~~3. w a ey. g.r %.~~
~~bid d M M 09i p A "s b'~~
~~u* '.a,4~~
~~- t 7 ~ i g M' a <Aaw c.' q ' g p'p' 1; uyjw%a e.,9 < ! r t 5 (o*s.'< F 6 d~~
~~m' D Ssm -,v.M~~
~~.%:,,.- ' 'p g e s' ?mF =- ,1 <~~
~~q}Q y u p~~
r. 1 t s i-N w,t 2 t m t es <@.q' e,+ a3:agg~ j n,.g . q V: = W ( yy ?*'q . m s '" ; o : ms Mt ' '!W r t s
j ~ GE Nuclebi Energy l ,23A4800 Resisjon F Class 1: - September 1990 :
[
- 23 A4800, Rev. I-Supplemental Reload Licensing Submittal . tor Pilgrim Nuclear Power Station : Reload: 7. Cycle 8 t - t .1: l l l f[f' .i Y . Approved: M - Approved: v. itj /J. S. ha ey, anager E. . ansen, '. Manager - Fuel Lie sin Reload: Nuclear ~ Engineering i~ l3 3, - i... .+,,m.
t Pilgrim-
~~awsm 7~~
Reload 7 Rev 3: i Important Notice' Regarding-' '4 Contents _of This Report Please' read J carefully' L i . This report was prepared by General Electric Company (GE) solely for Boston Edison i Company (BECo) for BLOo's use with the U. S. Nuclear Regulatory Commission (USNRC) for.- amending BECo's operating license of the Pilgrim Nuclear Power Station. The in' formation: contained in this report is believed by.GE to be an accurate and true representation of the facts : known, obtained or provided to GE at the time this rep l ort was prepared.- The only undertakings of GE respecting information i$ this document are contained in th'e contract between BECo and GE for fuel bundle fabrication and related services for Pilgrim Nuclear Power Station, and nothing' contained in this document shall be construed as changing said 1 contract. The use of this information except as defined by said contract, or for any pu: pose other - I than that for which it is intended, is not authorized; and with respect to any suc' unauthorized use, h neither GE nor any of the contributors to this document makes any representation or warranty (expressed or implied) as to the completeness, accuracy or usefulness of the information contained. in this document-or that such use of such information may not infringe privately owned rights; nor - 3 'I do they assume any responsibility for liability or damage of any kind which may result from such - ] use of such information'. r t l f i ( i h I Page 2 ; 3 i-2---
t / 6 4 Pilgrim: saus* : ' Reload 7 Rev 1 l i Acknowledgment ~ i The engineering and reload licensing analyses which form the technical basis 'of this Supple. mental Reload Licensing Submittal, were performed by E G Thacker II of the Fuel Engineering 1 1 Section. The Supplemental Reload Licensing Submittal was prepared by P. A. Lambert and' i verified by G G. Jones of Regulatory and Analysis Services, ~ .i r-l t 1 .1; b 'I 2 5 .1 ~ i Page 3 g
-l I-Pilgrim
~~zwsm -~~
1 Reload 7 Rev. I 1. Plant unique Items (1.0)* I-Appendix A: Increased Core Flow Throughout the Cycle. y 2. Reload Fuel Bundles (1.0 and 2.0) I Fuel Troe Cycle Loaded Number Irradi d P8DRB282 .5 24 - P8DRB265H 6 P8DRB282 6-112 BP8DRB282 7 32 P8DRB282 7 160 I New BP8DRB300 8 122' Total '580 i 3. Reference Core IAading Pattern (3.2.1) mwd /MT mwd /ST 'i Nominal previous cycle core average exposure at end of cycle:
~~18,362 16,658~~
) Minimum previous cycle core average exposure at end-of-cycle : ..18,362 16,658 from cold shutdown considerations: y ~ Assumed reload cycle core average exposure at end of cycle: 21,018_ ~ 19,067 l Core loading pattern: ' Figure 1 (
~~{ - ) refers to area of discussion in General Electric Standard Application for Reactor Fuel,'~~
4 '. NEDE 24011 P A 9, September 1988; a letter "S" preceding the number refers to the U.S. Sup _ plement, NEDE 24011 P A 9 US, September 1988. g-Page 4 L c y 4 r +
t l' i Ll Pilgrim zwsm l Reload 7' Rev. 1- / 4. CQculated Core Effective Multiplication and Control System Worth; No Volds,- 1
i.
' 20 C (3.2.4.1 and 3.2.4.2) l l - Beginning of Cpelei K,y,., Uncontrolled .1.1161' p Fully controlled 0.962 Strongest control ro'd out.. 0.990 R. Maximum increase in' cold core reactivity with exposure i into cycle, AK ' O.000z 5. Standby Liquid Control System Shu'tdown Capability (3.2.4.3) Boron - Shutdown Margin (AK) (PIm) (20'C. Xenon Free) L
~~675 0.040 6.~~
. Reload Unique GETAB AOO Analysis Initial Condition Parameters (S.2.3) Exposure: . BOC8 to BOC8+8767 mwd /MT (BOC8+7953 mwd /ST) ll Fuel Peakinc Factors ' Bundle Power '~ Bundle Flow' Initial Design 34tcal Radial Axial R-Factor (MWt) (1.000 lb/hr). MCPR BP/P8x8R 1.20 '1.62 ^ 1.40 1.051 5.454 108.6 1.44 L Exposure: BOC8+ 8767 mwd /MT (BOC8.+ 7953 mwd /ST) to EOC8 ' i l l-Fuel Peakinc Factors Bundle Power Bundle Flow Initial'- Design Local Radial AKial - R Factor (MWt) (1.000 lb/hr) - MCPR i BP/P8x8R 1.20 1.59 .1.40 1.051 5.355 ' 109.2, 1.47 7. Selected Margin Improvement Options _(S.5.1) Recirculation pump trip:. _ No Rod withdrawallimiter:
~~No Thermal power monitor:~~
No Improved scram time: No 'l Exposure dependent limits: Yes Exposure points analyzed: 2 L 1 b x .rg.s I
1 Pilgrimi zwsm : Reload 7 Rev 1-l 8. Operating Flexibility Options (S.5.2). Single loop operation: Yes ~ t Load line limit: Yes Extended load line limit: Yes i-Increased core flow: 1Yes Flow point analyzed: 107.5 % Feedwater temperature reduction: No-ARTS Program: . No - Maximum extended operating domain: No 9.. Core wide AOO Analysis Results (S.2.2) Methods used: GENESIS Nuclear with GEMINI ODYN and GEXL ~ACPR Flux O/A" Event (% NBR) (% NBR) BP/P8x8R Figure Exposure: BOC8 to BOC8+8767 mwd /MT(BOC8+7953 mwd /ST) Load rejection without 651 125 032 '2-4 bypass Feedwater controller - 663 132 037=- 3 failure Exposure: BOC8+8767 mwd /MT(BOC8+7953 mwd /ST) to EOC8-Load rejection without 730 126 035 4 bypass Feedwater controller 718 134 0.40 5 failure
=
'I l~ g.- Page 6 L
+ a ! Pilgrim; . mauo , Reload 7 ' Rev,1. -10. - Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary ] (S.2.2.1.5) ' Generic Bounding Analysis Results Rod B15ck ACPR. Beadin; ~ BP/P8x8R : 1 104 0.13 J 105 0.16 106 0.19-107 0.22 1 108 0.28-109-- 0.32-110 0.36 Setpoint selected: 110 y
~~11. Cycle MCPR Values (4.3.1 and S.2.2)~~
~~' Safety limit: 1.04 i 'Non nressurization events: Exposure range: BOC8 to EOC8 BP/P8x8R j Loss of feedwater heating ' 1.21~~
Fuelloading error 1.23*
Red withdrawal error 1.40 i Pressurization events: Option A - Option B BP/P8x8R : BP/P8x8R -t Exposure range: BOC8 to BOC8+8767 mwd /MT (BOC8< 7953 mwd /ST) Load rejectio'n without' 1,42 bypass L Feedwater controller 1.48 g
~~failure~~
~~The minimum MCPR value required by the ECCS analysis is 1.24.~~
l - " Option B is not available for this exposure; l g. .. Page 7 - l
y. j r Pilgrim : 2aaw - D ., Reload 7 ' Rev l' ' 11'. Cycle MCPR Values (4.3.1 and S.2.2) (continued)c A" - Pressurization events: I Ontion A - Ootion B 4 - BP/P8x8R -
~~BP/P8x8R-~~
,1 s ,y Exposure range: BOC8+8767 mwd /MT (BOC8+7953 mwd /ST) to EOC8: J Load rejection without 1.45f 1.40s 1 bypass j Feedwater controller 1.51 1.41-failure 1 12, Overpressurization Analysis Summary (S.3) PJ Pv 4 Event (D31g) .(gdg) Plant Response.- .MSIV closure (Dux scram)- .1317 .1333 Figure 6 i 1
~~13. Loading Error Results (S.2.2.3.7)~~~
~~\\ - Variable water gap misoriented bundle analysis: - Yes - o Event ACPR .j 5 Misoriented fuel bundle. 0.19 t~~
~~14. Control Rod Drop Analysis Results (S.2.2.3.1)~~
Pilgrim Nuclear Power Station is a banked position withdrawal sequence plant; therefore, the - control rod' drop analysis is not required, NRC approval.is documented in NEDE 24011 P A US, September 1988. i I j 1 l (' .Page 8 : .N
.c a ( l! Pilgrim i - nw,m - Reload 7 Rev 1 3 k'
~~15. Stability Analysis Results (S.2.4) i Pilgrim Nuclear Power Station is exempt from the current requirement'to submit a :~~
I cycle specific stability analysis as documented in the letter, C. O. Thomas (NRC) co H. C. Piefferlen (GE), Acceptance for Referencing of Licensing Topical Report NEDE.24011 Rev. 6,: ( Amendment 8,ThermalHydraulic StabilityAmendment to GESTAR 11,' April 24,1985. Pilgrim Nuclear Power Station recognizes the issuance of NRC Bulletin No. 88 07, = Supplement 1, Power Oscillatioru in Bolling Water Reactors (B1 yrs), and will comply with the y recommendations contained therein.
~~16. Less of coolant Accident Risults (S.2.5.2)~~
~~LOCA method used: SAFE /REFLOOD/ CHASTE See Loss of Coolant Accident Analysis Report for Pilgrim Nuclear Power Station, NEDO 21696, August 1977 (as amended). ?f i .k .i t i Ik [~~
4
..Page 9
s- - Pilgrim ! zwo:o - Reload 7 Rev. 1. MMMMMMM 1 E@@d@@$_E@+e@0+esad@00cs @_E k _@ 48 m esso o 4s M i+s M M M M M i+s M M i+
~~MMMMMMMMMMMM84 H~~
~~':MMMMMMMMMMMMM ':MMMMMMMMMMMME 1 ll:MMMMMMMMeMEME~~
~~MMMMMMMMMMMMM~~
~~'::MEMMMMEEM+8!MM24~~
~~2+EMMMMMs+sMMMMME4 MMMMMi+MMMMM 1~~
mdoo+oooo+ommemoomeoscE@@e@nnen e s 0+0@@@@0+0@@@Y@00 4 E@E00Ec0sOmE@i 2 I i 1: l l 'l. l l -l-I l. I l.l-1 3 5 7 9111315171921232527293133353739414345474951 FUEL TYPE j A = P80RB282 D = P80RB282-B =~P80RB265H E = BP8DRB282' C'= P80RB282 F = BP80RB300 8-Figure 1 Reference Core bading Pattern Page 10 ' 'j; 'i
+ . Pilgrim -. nuam Reload 7 ' Rev 1' j i 4 -- ?N?$hYrEcb'!"' !hfIhh L /; [ 3 \\\\ / E..,- g m.. y I N N, n /: . r.. t.. .n. n. .e - t sin cu:x.s> iiw cucacs> h;lkff'Nyjh6P A./ !'b IEENiiv p.& LE!v!!! afr il ssv / c ,y g a[. -s g&j a / 4 W n .e e t 9 \\ (. ~~s. s.. s%. .w. tiw (ucacs) tid (acmesi t t 1 ~ 1 i t., Figure 2' Plant Response to Lead Rejection without Bypass (BOC8 to BOC8+8767 mwd /MT [BOC8+7953 mwd /ST)) l -l Page 11 - ) i L.'
1 s' Pilgrim - zwsm - Reload 7 - Rev 1: i 1 i' r f.hakkI d I D r'a ,( a t rivi iS!!!?S ltL
l
~~!Li 15F. :P . ),k g =' a 1 I b f a-4 G -~~
~~L m: : :: l:-~~
~~e L . m. finit (Mtxt) f ad (Ecx5) iv gai [ g > i gua. val) '! & 5 ! ' M } i Ir _'C i bf _e. - 2 2 E 2* r. r. y -g-~~
~~l_U'y, L-~~
'/G, ( -l -l r u... q iin (ucxs> iin (ac.csi / ~i: Figure 3 Plant Resp'onse to Feedwater Controller Failure (BOC8: to BOC8+8767 mwd /MT [BOC8+7953 mwd /ST)) t' i Page 12 Ig- -L. mm i +
8 Pilgrim tesm Reload 7 Rev 1 t 0 d 's$r'#r,ot roui Q'WIIOEf) ' c 4 mtf ooy > h!U v4 Dh a = a MmN / i g =.- sx 7 -l N =.. a ._,',m
~~. =,.~~
~~..v. - -- n.- 3.- 1 Ild (ECO@$) Ild (EC.CS)- 3 J/ega [ viiv -g i'IU,," - 3M{v6i,fif[T5EE5681) gRgA* 1L lo S )1 .f e 1m S jA A E -f \\ x t ") v - Vv~~
~~i 5~~
~~k .e u .g~~
~~1: :' :~~
I 1 k = =.. 84. l. ud cuc cs> lid inc.es> l. L l-8 Figure 4 Plant Response to Generator Imad Rejection without Bypass l (BOC8 +8767 mwd /MT.[BOC8 +7953 mwd /ST) to EOC8) ' ' Page 13 l y.
I Pilgrim. zwsm ' Reload 7 Rev 1: -a j i t g I .a URTa h[ai rLys y . I '!h.d T'f? h# gik k rdl k' ter... 1 c-A C...;g,e \\ g t t ' i f f m \\ ~ t 1 .i s i ) ,m .l .m. a. i ' tid (MCats) ~ lid (utecs) --. i (i p.9 RT) ivi aclidhi I L{vI' gn }.R r.$ ' d' Ef f I I f.^ _ r . I. /s _= g' > -_7 . y q n g i f 'l ^ as. .i.. 4 '.1 s e... _ tid (u:acs) tid (ucoes) - l .L i 0 F!gure 5 Plant Response to Feedwater Contmiler Failure (BOC8 +8767 mwd /MT [BOC8+7953 mwd /ST) to EOC8) Page 14 : J% o )
~~8. -~~
~~Pilgrim ~ mam Reload 7 Rev l' t i 1e 14 5. F : 1 LfP(!) ((Pil) i $dRfA~~
~~(al fWx 114v v~~
~~I I INL[ fLGs {f%4W l{ w 'i ( " m. I~ 1, _. m %s c b i 4 b ~ r ~ \\ s a a 'l i /, ~f-~~
l
. m s' er. i. ~ tid ($CC405) tid (5tcecs)- i i so:. .sp. sat) .i viuiv 5 Ow 4:livait .jov - glp-a r =. . i. 9' A A s..:. =.. 4 z lv4 M,.. h... k 11 r ~r. n.. 3 lid (ECDCS) tid (gcess)c I 'f t t 'E Figure 6 : Plant Response to MSIV Closure, Flux Scram t t Page 15
i g' : Pilgrim 2wam ' Reload 7 Rev 1 Appendix A'. . Increased Core Flow Throughout Cycle 8: The analyses performed for C cle.8 included increased core flow throughout the cycle. 1 There are no concerns regarding reactor internals pressure drop of flow induced vibration' as . discussed in the increased core flow analysis document for the EOC 6 (NEDO 30242). l .i a The flow biased instrumentation for the rod block monitor should be signal clipped for a setpoint of 107% since flow rates higher than rated would otherwise result in a ACPR higher than. reported for the rod withdrawal error.
p.
( ci I .i 1 ll d l Page 16 - (Final) p l' ~.
o g I, 0CFR50.90= 1 1 sormsmorsou Pdgom Nuclear Power statton' Ro:ky HJ! Road ( ?tymouth, Massachusetts 02360 : . L Ralph G. Bird senior vice President - Nuclear BECo_ 90- 136 - U.S. Nuclear Regulatory Commission-Document Control Desk-Washington, DC '20555 J 0l License DPR Docket 50-293 I PROPOSED TECHNICAL' SPECIFICATION. CHANGE MINIMUM CRITICAL POWER RATIO-Boston' Edison Company-proposesithe' attached' revisions to-Appendix A ofi l Operating License DPR-35 for the Pilgrim Nuclear.. Power Station in accordanct j with 10CFR50.90. The proposed revision' toithe Technical Specifications' upgrades the Safety Limit Minimum Critical-Power Ratio and revises-the L Operating-Limit Minimum Critical Power' Ratio. These proposed revisions.will extend-the use of spectral shift to maximize fuel utilization during the_present fuel cycle. This change is not needed if, our proposed change of August 21, 1990 is_ approved first, .]-' h[( <^ , j e R.- rd
~~.; p RAH /njm/4820 1~~
Commonwealth of Massachusetts) 1 County of Plymouth ) ^ Then personally appeared before~me, George H.-Davis, who being duly sworn, did-state-that he is Vice. President - Nuclear Administration of. Boston Edison-Company and that he is duly authorized to execute,and-file-the submittal 1 contained herein in the name' and'on' behalf of Boston E'ilson. Company and: that the statements in said submittal are true'to the best'of his knowledge-and belief. My commission. expires: O M d'/995~ d ## I u DATE ' ]' NOTARYP0flC-l Attachments: A. Description of Proposed Changes B. Replacement Technical Specification Pages i C. Marked-up Technical Specification-Pages-D. Supplemental Reload Licensing Submitted for Pilgrim Nuclear Power Station: reload 7 Cycle 8, 23A4800 Rev. I j i signed original and 37 copies -cc: See next page __)
t BOSTON EDISDN._ COMPANY + U.S. Nuclear Regulatory _ Commission: Page 2 e cc: Mr. R. Eaton, Project Manager. Division of-Reactor Projects - I/II Office of Nuclear Reactor Regulation > x sMail Stop: 14D1. U. S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockville Pike Rockville MD 20852' i U. S. Nuclear Regulhtory Commission Region I_ 475 Allendale Road King of Prussia,1 PA 19406 Senior NRC Resident Inspector j ' Pilgrim Nuclear _ Power: Station: -j e Mr.-Robert M. Hallisey, Director; Radiation Control trogram Massachusetts-Department of Public Health = 150 Tremont: Street, 2nd Floor i Boston, MA 02111 -) i .l 't j l l h
Attachment A to BEco 90-136 Description of Pronosed_Chanat Pronosed Chanagi Boston Edison Company propose to upgrade the Safety Limit Minimum Critical Power Ratio (MCPR) and revise the Operating Limit Minimum Critical Power Ratio. The revision to the Safety Limit Minimum Critical Pow 3r Ratio reflects improved fuel designs in the core and the General Electric Boiling Water Reactor Generic Reload Fuel Application, NEDE-24011-P. The revision to the Operating Limit Minimum Critical Power Ratio is being proposed to be consistent with the upgraded Safety Limit MCPR and to reflect spectral shift operation. Spectral shift operation results in top peaked flux distribution patterns toward the end of the fuel cycle. A top peaked flux distribution pattern increases the change in the Critical Power. Ratio (ACPR) associated with the analyzed operational transients. The proposed revision maintains conservative operating limits. The upgraded Safety Limit MCPR and the revised Operating Limit MCPR ensure the plant will be operated safely ar,d will not pose an undue risk to the health and-safety of the public. These revisions will allow the plant to be operated more efficiently. Basis for Chanae The NRC app oved the upgraded Safety Limit MCPR in its safety evaluation for Amendment 14 to NEDE-240ll-P-A, " General Electric Standard Application for Reactor fuel", dated December 27, 1987. Inco.*poration of the upgraded Safety Limit MCPR is appropriate for FR's with D-lattice fuel assemblies provided: 1) the fuel has a beginning on life R-factor greater than or equal to 1.04 and consists of fuel types P8x8R, BP8x8R, GE8x8E or GE8x8EB; 2) the fuel is at least 2.80 weight percent U-235 bundle average enrichment; and 3).the lower enrichment bundles in the core have operated for at least 2 cycles. The fuel asseinblies presently in the core of the Pilgrim Nuclear Power Station satisfy the conditions evaluated in NEDE 240ll-P-A. These conditions will also be satisfied for future reloads. The Supplemental Reload Licensing Submittal for Pilgrim Nuclear Power Station, Reload 7 Cycle 8 submitted by BECo letter 87.081 dated May 22, 1987 has been revised to reflect operation using a spectral shift fuel management strategy. The revised analysis also reflects an assumption that the Turbine Bypass Valves do not open in the analysis of the feedwater Controller Failure with maximum demand. This assumption was mhde because precise measurements of the oper.ing time are not available. The assumption is conse vative and it results in a more severe transient analysis and a more conservative Operating Limit MCPR. This limit is incorporated in the proposed Technical Specification. The proposed Operating Limit MCPR will ensure the HCPR does not decrease below the i proposed Safety Limit MCPR at any time during any abnormal operating transient i as defined in the FSAR. i
l Determination of No Sionificant Hazards Considerations The code of Federal Regulations (10CFR50.91) requires licensees requesting an 's amendment to provide an analy.)s, using the standards in 10CFR50.92, that determines whether a significant hazard consideration exists. The following analysis is provided in accordance with 10CFR50.91 and 10CFR50.92 for the proposed amendment to Pilgrim's Minimum Critical Power Ratio. 1. Upgraded Safety Limit MCPR A. The proposed change does not involve a significant increase in the { probability or consequences of an accident previously evaluated because the proposed change to the Safety Limit MCPR does not change plant equipment, operating procedures, or governing design criteria used to protect the 3 plant against the initiation of any analyzed accident or used to mitigate l the consequences of any analyzed accident. B. The proposed change does not create the possibility of a new or differnt ( kind of accident from any previously analyzed because the proposed change { does not change plant equipment, operating procedures, or governing design criteria and the change to the Safety Limit MCPR provides the same level of protection as the existing Safety Limit MCPR against fuel cladding failure during an abnormal operational transient. The proposed Safety Limit MCPR 4 therefore provides equal assurance against a release of radioactive I material in excess of 10 CFR 20 limits during abnormal operational transients and a new event sequence leading to an accident is not created. C. The following design requirement ensures an adequate safety margin is maintained: Abnormal operational tralisients caused by a sinf e operator error or equipment malfunction shall be limited such that, considering J uncertainties in manufacturing and monitoring the core operating state, more than 99.9% of the fuel rods would be expected to avoid boiling transition. The proposed change does not involve a significant reduction in the margin of safety because this design requirement, which governs fuel cladding integrity and maintains the defense-in-depth philosophy, has not changed. 2. Revised Operating Limited MCPR A, The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the proposed change to the Operating Limit MCPR does not change plant equipment, operating procedures, or governing design criteria used to i protect the plant against the initiation of any analyzed accident or used i to mitigate the consequences of any analyzed accident. l
i B. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed change does not change plant equipment, operating procedures, or governing design criteria and the changes to Operating Limit MCPRs provide the same level of assurance that the Safety Limit MCPR will not be exceeded during an abnormal operational transient thereby assuring a release of radioactive material in excess of 10 CFR 20 limits will not occur during abnormal operational transients and a new event sequence leading to an accident has not been created. C. The proposed change does not involve a significant reduction in the margin of safety because the conservative Operating Limit MCPR ensures the most i limiting transient will not violate the.$4fety Limit MCPR. j Reauested Schedule The requested change is not needed until the present fuel burnup reaches 7200 MWD /T, about January 31, 1991. Additionally. approve
~~of this request will not be needed if our proposed change of August 21, 1990 is approved first.~~
l h l 'l )
i Attachment B to BECo 90-136 List of Effective Pages i I Revised Pages i e 6 12 c i 13 2058-2 t L i I i i k I ) f I
s - 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING j 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRIT_Y l Aeolicability: Anolicabilitv. j Applies to the interrelated-Applies to trip ~ settings of the variables associated with fuel instruments and devices which are i thermal behavior. provided to prevent the reactor i system safety limits from being Obiettive: exceeded. 1 To establish limits 'below which Obiettive: l the integrity of the fuel clad-4 ding is preserved. To define the level of the process variables at,which automatic i Soecification: protective action is initiated to. I prevent the fuel cladding i A. Reactor Pressure >B00 osia and. integrity safety limits from being i Core Flow >10% of Rated
~~exceeded, i~~
The existence of a minimum Soecification: critical power ratio (MCPR) less { than 1.04 shall constitute A. Neutron Flux Scram-violation of the fuel cladding e integrity safety limit. A MCPR l of 1.04 is hereinafter referred The limiting safety system trip to as the Safety Limit MCPR. settings shall be as specified below: B. Core Thermal Power Limit (Reactor Pressure 1800 psia and/or Core
~~1. Neutron Flux Trio Settinos Flow 110%)~~
When the reactor pressure is ' a. APRM Flux Scram Trio l Settina (Run Mode) I 800 psia or core flow is less than or equal to'10% of rated. When the Mode Switch is in the steady state core thermal the RUN position, the APRM power shall not exceed 25% of ' flux scram trip setting design thermal power. shall be: C. Power Transient S 1.58H + 62% 2 loon The safety limit shall be assumed Where: to be exceeded when scram is known to have been accomplished S = Setting in percent of by a means-other than the .ratedtthermal power expected scram signal unless (1998 MHt) i analyses demonstrate that the j fuel cladding integrity safety. H = Percent of drive flow 1 limits defined in Specifications to produce a rated core 1.1A and 1.1B were not exceeded flow of 69 M lb/hr. during the actual transient.- i i l Revision 6 Amendment No. 72 f
RMES: The statistical analysis used to determine the MCPR safety' limit.is based on a model of the BWR core which simulates the process computer i function. The reactor core selected for these analyses was a large 764 assembly, 251 inch reload core. Results from the large reload core analysis apply for all operating reactors for all reload cycles, including equilibrium cycles. Random Monte Carlo selections of all operating parameters based on the uncertainty ranges of manufacturing tolerances, uncertainties in measurement of core operating parameters, calculational uncertainties, and statistical uncertainty associated with the critical power correlations are imposed upon the analytical-representation of the' core and the resulting= bundle critical power ratios. Details of this statistical analysis are presented in Reference i 2. - 3
~~8. Core Thermal Power Limit (Reactor Pressure < 800 esia or Core Flow~~
) < 10% of Rated) The use of the GEXL correlation is not valid for the critical power calculations at pressures below 800 psig or core flows less than 10% of rated. Therefore, the fuel cladding integrity safety-limit is established by other means. This is done by establishing a limiting condition of core-thermal power operation with the following' basis. Since the pressure drop in the b" pas L region is essentially all elevation head which is 4.56 p' the core pressure' drop at, low power and all flows will Iways be great than 4.56 psi. Analyses show that with a flow of 28x10 lbs/hr bund 16 ilow, bundle pressure drop is.nearly >i independent of bundle power and has a value of 3.5 psi.= Thus, the bundle flow with a 4.56' psi driving head will be greater than 28x103 lbs/hr irrespective of total. core flow and independent of bundle' power for the range of bundle powers of concern. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate'that the fuel. assembly-critical power at this flow is approximately 3.35 MHt. With the design peaking factors the 3.35 MHt bundle power corresponds to a core thermal power of more than 50%. Therefore a core thermal power limit of 25% for reactor pressures below 800 psia, or core = flow less i l than 10% is conservative. 1 l l-Amendment No. 42 12 e .P
JLMLS: \\ C. Power Transient Plant safety analyses have shown that the scrams caused by exceeding i any safety setting will assure that the Safety Limit of Specification 1.1A or 1.1B will not be exceeded. Scram times are checked j periodically to assure the insertion times are adequate. The thermal l power transient resulting when a scram is accomplished other than by j the expected scram signal (e.g., scram'from neutron flux following 'j closures of the main turbine stop valves)'does not necessarily cause fuel damage. However, for this specification a Safety Limit violation will be assumed when a scram is only accomplished by means of a backup i feature of the plant design. The concept of not approaching a Safety { Limit provided scram signals are operable is supported by the extensive plant safety analysis. ] The computer provided with Pilgrim Unit I has a sequence annunciation i program which will indicate the sequence in which events such as scram,- l APRM trip initiation, pressure scram initiation..etc., occur. This program also indicates when the scram setpoint is cleared. This will provide information on how long a scram condition exists and thus provide some measure of the energy.added during a transient. D. Reactor Water level (Shutdown Condition) During periods when the reactor is shutdown, consideration must also be given to water level requirements due to the effect of decay heat. If i reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core can be cooled sufficiently should the water level be reduced to two-thirds the core height. -Establishirent of the safety limit at 12 inches above the top of the fuel provides adequate margin. This level will.be continuously monitored. Reference 1
~~1. General Electric Thermal Analysis Basis (GETAB):~~
~~Data, Correlation and Design Apolication, General Electric Co. BWR Systems Department, November 1973 (NEDO-10958).~~
~~2. General Electric Boiling i4ater Reactor Generic Reload Fuel Application, i~~
NEDE-24Dil-A. i Revision 1 Amendment No, 42 13 lo
i i t TABLE 3.11-1 OPERATING LIMIT HCPR VALUES A. MCPR Operating Limit from Beginning of Cycle (BOC) to BOC + 7,953 MHD/ST. P8xBR/BP8xBR l For all values of t 1.48 B. MCPR Operating Limit from BOC + 7,953 MHD/ST to End of Cycle. pbx 8R/BP8x8R t f t10 1.41 0.0 < t 1 0.1 1.42 0.1 < t 1 0.2 1,43 0.2 < t 10.'3 1.44 0.3 < t 1 0.4 1.45 0.4 < t 1 0.5 .1.46 0.5 < t 1 0.6 1.47 0.6 < t 1 0.7 1.48 0.7 < t 1 0.8 1.49. 0.8 < t 1 0.9 1.50 0.9 < t i 1.0 1.51 i i L Revision L Amendment No. 708 205B-2 l
i 1 ATTACHMENT C to BECo 90-136 'i Marked UD Technical Soetifications } t 5 I-t 14-1 k ? l
t i 1.1 SATETY LIMIT 2.1 LIMITING SATETY SYSTEM SETTING l2.1 TUEL CLADDING INTEGRITY i 1.1 TUEL CLADDING INTEGRITY Appliesbility1 l Applicability: I Applies to the interrelated Applies to trip settings of the variables associated with fuel j instruments and devices which are thermal behavior, j provided to prevent the reactor system safety limits from being i exceeded. l t Objectives Objectives 2 e To establish limits below which To define the level of the pro-l cess variables at which automatic the integrity of the fuel clad-ding is preserved. I protective action is initiated to prevent the fuel cladding inte- _grity safety limits from being l exceeded. l Speciffratient Specification: A. Resetor _ Pressure >800 psia and A. Neutron Flux Scram i Core flow >101 of kated The existence of a minimum i= The limiting safety system t rip critical poser ratio (MCPR) settings shall be as specified /, g less than, shall consti-below: tute violation of the fuel cladding integrity ety-Neutron Flux frip Settings-limit. A MCPR of is here-g inafter referred to as the' s. APRM Flux Scram Trip Safety Limit MCPRi t Settina (Run Mode) .l 6 B. Core Thermal Power Limit (Rear-When the Mode Switch is i tor Pressure 5800 psia and/or f in.the RUN position, ~ the APRM flux scram Core Flow 110*d = trip setting shall be: 5 800 psia or core' flow is less-S S.58W + 62% 2 loop. f When the reactor pressure is q_ than or equal to 10% of rated, the steady state core thermal ( Where power shall not exceed 25% of j S= Setting in percent design thermal p ver. 3 of rated thermal C. Power Transient l power (1998 MWt) I 'W= Percent of drive-The safety limit shall be as-' sumed to be exceeded when scram flow to produce a is known to have been accomplished, rated core flow of 69 M lb/hr. 't by a means other than the expected scram signal unless analyses demon-strate that the fuel cladding ? integrity safety limits defined in-Specifications 1.lA and 1.1B were I a not exceeded during the actual, ] i transient. 6 Amendment:No /[g l-a ~
~
~~A R /ABEA 7~ "A " r O. r e ..e.yo ; t., the.t.c.;1rel ~;e1 . we e.i t.1.cn / l s -ide;ed n tem *4-1r uf::: n 3. de r d-~~
~~nl=: !! 9: n::~~
~~y i / punnn: 14+ted - h i:bl+-M. Ref e.-ee 3..o; e.. 1.dr. i eeesebly pan die;siLnie; ehe-Z 71,.see bl ? >1. vi~~
~~ef... <..* -,ee n s u. : > =, 1.f n = 4, 2 = 1~~
~~-fe;;;; distribett = - P-t n: ' 7 : t *2 -l-et ti"" "' " ..... ~..... ,=.w. .u. ..uv. .u. -distenbett::: h' n; nhn ;27 th h;;i ';;; ;f it h'.'~~
~~..1;.~~
Th hei -!+epS.uneeet '--i:: 1: S: n;; prznr: n; 7 :n l i: M OO 10360 rd O. h;1; fu O.;.;.;en ie;j ia ;; e W eeeeel* sten-6e-3 *en-4a-h S 10?$$ O.; ;xn din;it.:in i 6 4; u nd-;;-e-typteel 41 2:- -117 n:: 1: tid $2 ::d ;:ttir_ e; erli;;;;1-ly-eheses-te pren;; ; ;h..;d pan din-ii.%;; 1.'... -;h g:n:eet-cri:: ;f-n : M 11e; n O.; highen p;..; 1.4 b. T- - u n: diese4 bet &em-4a-Mig-1: rnin 7s n M: tic: 'Jd: 1 dri ;
7 N

1-eyele-would n ; M.; seveie I.L. 41.i.i'uv uvu..a 4.
the nelj.;;c 3. Cere n ormal power tir1: (Reseter pressure < 800 psir er Core T1ov < 10: ef F.ated)_ l De use of the CEII. cerralation is not valid for the critical power calculations at pressures belov 800 psig or core flows less than 10% of rated, n erefort,3the fuel cladding integrity saiery id*' t is established by other means. This is done by establishing a limiting condition of core thermal power operation with the folleving basis. Since the pras'sure drop in the bypass region is essentially all elevation head which is 4.56 psi the core pressure drop at low power and all flows vill always be greater than 4.56 pai. Analyses show that with a flow of 2Ex103 lbs/hr bundle flow, bundle pressure drop is nearly independant of bundle power and driving head v'11 be graatar than 12x10ge flow s-ith a 4.56 ps1 has a value of 3.5 psi, nus, the bund lbs/hr 1r espective of total core flow and independant of bundle power for the range of bundle powers of concern. yull acale A".uS test data takan at pressures from 14.7 psia to 800 psia indicate that the foal assenbly critical power at this flow is orproximately 3.35 MWt. With the design peaking f actors,the 3.35 MWt bundle power cer-responds to s core char =al power of more than 50%. Theref ore a core thermal power 11:it of 151 fer raatter pressures belov 800 psia, or core flov less than 10: is conservativa. \\ l A::endment No. 11 w (
Insert Y The statistical analysis used to determine the MCPR safety limit is based on a model of the BWR core which simulates the process computer function. The reactor core selected for these analyses was a large 94 assembly, 251 inch reload core. Results from the large reload cor& analysis apply for all operating reactors for all reload cycles, including equilibrium cycles. Random Monte Carlo selections of all operating parameters based on the uncertainty ranges of manufacturing tolerances, uncertainties in measurement of core operating parameters, calculational uncertainties, and statistical uncertainty associated with the critical power correlations are imposed upon the analytical representation of the core and the resulting bundle critical power ratios. Details of this statistical analysis are presented in Reference 2. d
...:.;;.~;;; ~ r-- o J i I C. Pever ransient_ Plant uf ety analyses have shown that the scrams cauud by ex-i caeding any saf ety settint vill assure that the Safety Limit' of specification 1.1A or 1.3 vill not be exceeded. Scram times are checked periodica2.17 tu assure the insertion tfJbes are adequate. The therrel power traasient resulting when a scram - is accomplished other than by the expected scram signal (e.g., scram from neutron flux fon oving closures of the main turbine stop valves) does not necessarily cause fuel damage. Bowever, f or this specification a saf ety Limit violation vin be assumed. when a scran is_only accomplished by means of.a backup-fascura-of the plant design. The concept of not _ approaching a saf ety Lir.it provided ' scram signals are operable is supported by the extensive plant safety analysis. The computer previded with Pilgrim Unit 1 has a sequence annunciation program which will indicate the sequence in vhich events such as scras, AFRM trip initiation, pressure scram initiation, etc. occur. This pro 5 ram tiso fadicates when the scram serpoint is cleared. This viu provide information-on hov long a scram condition exists and thus provide some measure of the energy added during a transient. Reacter k*ater Level (Shutdown Condition) D. During periods when the reactor is shutdown, consideration mast also be given to vster level requirements due to the affact-If reactor veter level should drop' below the of decay heat. top of the active fuel during this time, the ability to ceal tne core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core can be cooled 'sulficiently should-the water level be reduced to tso-thirds the core height. Establishment of the saf ety at 12 inches above the top cf the fuel provides adequate 11=1 This level viu be continuously monf.tored. marsin. Ref erences_ General Electric Thermal Analysis 5aris (GETAB) - Data, 1, Correlation and Design Application, (aneral Electric Co. 3WR Systems Department, Novembs.r 1973 (NDO-10958). - -- W y-1.' n :::;; 0;g ut: _7: f: n- :: 1-12 ti' r i^ 22 ", 0- eral-41;;tri: 0 rprry 5'a Syctr- 0 perrt, 220, 1SM (=0 - 20!!O). / General Electric boiling V;ter Reactor Generic Reload A Tuel Application, ECI-24L.1-?. .eeneentso.g u ~ p 1 ~ ~. -..1 .... i. .~
TABLE 3.11-1 OPERATING LIMIT MCPR VALUES A. MCPR Operating Limit from Beginning of Cycle (BOC) to BOC + MWD /ST. PBxBR/BP8xBR 7,953 D/.46 For all values of i / B. MCPR Operating Limit from BOC + MHD/ST to End of Cycle. 1 PaxBR/BPBx8R / 11 0 /.M/ h/,#2 O.0 <11 0.1 0.1 <t1 0.2 h/.M3 h/.MY O.2 <1 1 0.3 0.3 < t1 0.4 /,45 0.4 <t1 0.5 4,40- /* N (/ f 0.5 < t1 0.6 -h40- /M7 0.6 < t1 0.7 / YO 0,7 <t1 0.8 +:43 /, Mf [/ s O.8 <t1 0.9 h44- /.60 f 0.9 <ti 1.0 %44- /, f/ / i / f i l i Revisionh 205B-2 u d /le. o l r
m, q:w a,. w, w t.g. ss si g W .. < e4 m o 4 W(:f,iQm,]f f f, y* b, }.f}m.:L @ }m o = &, ', i}7-[
~~jy m~~
~~_ * [,. p g Q, nQg_,~~
~~f g, w s g;-~~
a 4 w'.. ., N l r w,.e.c .-. - i. ~p' YfA + a 'Q. .- b *q, 1 41 -s g =_a>, g f. w(l ' ' N:f ' y- - Q4s G:Q&'._W \\- NQ* p ' e u ,ra. ' Q,M ; .OL . - y Gd.. QQ - %W4 ~. h, %;. )e a? hW ' d, Q{lKl' Q. ,? &l'% iMO^ b% (uQ, ? /J tQ(y,hY* Qi +% J t ? , j.:. b,
~~M l~~
~~.r .s', g .w. y ; :.s u s ' 4< t o : u n g', -u w. m ', ? - 5 in g, y.. 1, o ' p i '.,r ^ g; q *, r' s, +A', >'!4. ;"_ %~~
~~wt.~~
.a# T eu ',.. ; s i r 1' r g ~ g 9 x h ii ' =-,:a 'd 3 y. '. S iM.?U h' f[1 P' ""9 t d f. 4, ; r t,. _k ; . -i- , M ; :. u t-3.- I t ag 3., y;WLp r y 3,y a_ g[.., ,) p s; { ); f-- ) 4 "? E c 23.< .t i y . 3 g,,a W," '. W ' t r w ,y /; ;,g. 4 g,*l + + h p ~,,' ? L.' L_l T m. 1 ....,f ?, 3~
~~r,,l t~~
w 9 c w 5 m_ < 7 ' - o(g?? 4t a, s ,. /,'.a-, w 1 1 ' b 6"" Vi-np -t } ~y i .J, 1 r, -, r m' 3 $n'.,,t, +c 4 ' ATTAC,HMENT; D t o MCo 90-13,t7 s t =i o m,y ,.w m q _ - - ,j ,S,,- L t s ;.n - a 4 s g; g., ( s M. .Av 4 the/ 's ..., --> s so (y ' Supplemental 1ReloadLicensiAg,t-$ubmitted,Jor Pilgrim NuclearfPower-7 W ** j;,m.g, j e.
~~m. -~~
2 Station reload % Cycle 8 d23A4800fRev M 1. & ' Y,7, s1h' ,.. ; j' s a -[-, +- 6 4 ~ g.3 4 -e = i ' - < r ,y _ N. e_.. 5 .( 0 .w g, >_ i 2 i. I i i 4 t -4 i t s - r, . li q i P i f pp 1 3 As f ,' c.W J in i n 5 j.; u, ,3.f t s ~y'#
~~2qm4 ',~~
s t-4 c7-m s g-.., s) ', s-,, g s. =.,- eu > c c 1 n. G.-. i+g:i u , w. t ': ,4>g 1 f 1 4 .ig .y ' ' -{.,. ,p,, g sF y} Q., ^ b,z 1.i.m .pi! t = >;i, yy9-y> j g' 7, n' t[ - (g; g %g s v s ., s. b gy+ qq, . i d p.q ^ a , - i .. [p ; q, 4 + 3. g j l. y '.t, ,,t y 1 8 1 4 t l f-g- . gie i' 3 ' - M ,. >o r 9,) g t g p 9 A) 3' s' +.> w -k 1 n: t V a, _.m: , rw fi ; - ,: p[/ y m n p a ., [,. "'. + > s ,,, j ' ' 'yd f]' @ ' }m y \\.S $a, y - s h f(4" I so ~ v,Q.j g' s -, c j gjgyl 2+,,g f f * ,g qu l-f p s s 3 . + s_. eg rG 2.n ? * ' = rt ..v e Q o s y , u%.._, , umi 8 ,, _ p' s.' _ J a;[ ,ll - Q... SQ-t c '),_. '^ ,:Q s po if % f r n, v%?,.#f
~~y. f 5 7,~~
,3
~~q.. -~~
~~ajs. u.7 4r;1En ~ t - of / ge - <;rd c .9.. -- p; a n: 4dk ? f. - . o ,M<p W* n + N 4 .=3,- 2 v m -y s. r w+ ' y' 5 f.4, 2 8' r' s '. ~ t-~~
~~f 3t a,~~
9n. --q,< ,b,... '".<.E L }. =,_ 7i a4u . i. 1 .. e. c .;y ;r 3 M k' d'. a. f i} 3 k. Mh h ,e lh ,k 'l 0D f 7 ' fpl %A nW:?. {D-l w ns; +> y i . jfl 4 oW'- , e 1 s eW, n.r L' L;_., 'lb,+ Q.. G:: Q-,.,_, .t a ,J. i ( =.h ',^ l li. ,\\ } i 1 c k Nt',,, , r. f. < 7 } = ' o m ..s s a '
~~k, x~~
~~'{ ,'p* 5 ...g,y -'4 f,. 3.= M[ ! '-' M.l..t a T M' --y. ' + g ),3,c}-.J ' 6.j p L -[ 5: v "',t* -j ? F ![i g,, a,, > =. " ' P*'a'< 4 L g k ' k.- 4 1. "]< ' 9' 1..- t e dN <c ~qJy s t_q 1~~
~~- lN l~~
~~5 t n ? -' p' t -n~~
~~m..,~~
~~re-, .,>g r q.1 t 4 1 f^ -,u 4 2 ,p . ; f. P j a, 4 s 4 ,( - -3 r .b.1 + W' w-s 3 y _f- -y .,-p I' n ,rs f + } ;;;'. fr f,~~
~~%I s~~
4 =- -: 4 y : J ~, y l .y t a.- y h_, '1; 1 1[ 2h h N. P 3 4 _I. ( g, p' "h: 5 i 3' 'r r . J. -Y A .!s / k q , ; =. g ' s s m-e s 3 A*/ T l-1 r 4 g3 : '. }e;' 3l L <ra Y n-t.qj' [ f! h ' h.d h . = -f g '.{ tg' t' y g g yg'i'( s p e - g :(I,,, j ',,N,, s a 3 lT m;; i.:R,.. i > i,,m y ,4 ,s.., 1, , ( y. .p-t A-an 2 r I t'L# $ 1. [ f I. ~ g ,.y 1 m 4 .,h et, - t' u, , = i - p- ?- y - t 2 - -,cch - 4 g. d.' -? /{z ' Q, 4' ' O, _l.y } 3 t c e..... e 'l4 Vx , Q,> : .h F 2 4 @k h? , q y + ~ s 1c t q+r 8 e-s ,, v ?, {. -_h i y 3 4% 'Tm
~~y -~~
~~-y: y' 5 -) ,pf w pl:,,, 3 i~~
~~4. p-4 gj~~
~~.g) 4; 2 . f. aA s._ ge > > _n. + / rq~~
~~w ;,~~
~~,,7 ; .J, s. r 3 g a r.. ,4 s.s , ha c - 4 1 i , y] n-r s y y a: k,- j _, T ]" V., p c. O D, t~~
)
~~i n i 1 g 9 np a - n. yd ' ', ;l f b s, 1 g, -y o c 4 m O .e peg i , G vg - M_ i g-.. '-90< u, u &,, 4j ,+: .4 4 1 thi. 4 , t, i 3 '.'( t. 5 Ii; 7 5 c! i-y,, g -t~~
~~y-l-4%~~
~~m v i. w g-m y 5 .h" n' - n. , Lj ep ...) j' j' pi y v }li }' j 'bI. I' u r i J g~~
g'
~~+ u t c t S,g Y < -(_ [. :, 6 !A 2 3.p] v g" -l ' c c .)-" 3 s 1~~
[
_? st i 't g 5 m 5 r. b k' L' . f'sr I 9. g' jg T f T 'T w E J 1 ^b I g .-s 4 -l ,3.; h - 3 4 a y cn' i N ,f N 5 ._h{ .,. -.$aa 3 , 1 u,y s q,i. _. ..u..; - c. ,,l..s .1o t t v;p;h r G [', .- ' _V ,g. [. 9 , )% \\ p m. lm,-..' rt j . ;.-n-f", Ik j 4 f_-..-_ -_g 3., i y 4 ^ : ,4 , !d N $.? . f g _ p. ;.lV 'f* - u M ip.2 K " OnS: i 9 -y -' 4'j'. 'Q* '4, ,,6 7 iaf ).r. = Q :Q + -Mk ^ _q:, y-2 t i. u r h f 97 9 : y g m' t-g } .g. z.' '.gj,;.j. t m O n 4,3 M{.y 3- /M jp 39 n. ..s 30 t uC ii,. . sus g 'y t 1),5' yf (, g ( "-h P, c^ o ( f 4 l,. v ' L ) ~ _j h.' T ,i. p a". i .k 6 m.# i I 3 , s. - . i i, . g (, p',.[t'}.- -,g, 5'h D r. 1 tt b.I.T.O, y (,h f. ~.a, a e-h I t'_ Tg- -P gx>* w a q j 4 t.g -. m .y -s g ]y -lWC m 2# m, R t. A M 1 - 7 + g, =. ap q _. sh ,1.f.ps ; ast v m 4 )-, t n [, v n.-,[hi !i yb' 4 4 ' 4 4 d- [,2 a. h 3,( 5' %.m A ,,. i. ^ r nw n , p X'l 3- 'Q[ e-t:- ~' Gh r _,WWL / 14 M n.+y?, 5 ,1., j, ' 3W-m p snB %g,?..g e t a ,4, a~,%'i t s v w wc
g
~~) gy vy y ,' w gtw i .;j g ^ 8 . l s r 4. g gt M I 1 c n* wh Mau + j A y!~~
~~$9 ' '. y, W N Q.W V}b n>~~
~~.:np 2 "d' . d 'O - /pf Wa-mm &.d, 'P'; i i S.p[7 ins W ff - F.4i4~~
~~1. g ( O u... >~~
~~e t~~
~~A- ~ -~~
M .i mm As y mM aw o M i,'. y ' m wmsm_.e . e w-s.%. ti _; g.,. =jY f IN?I .- n' , b g % %,'- gh, h.9 .,, m.h':. [ .)'- f+ h b s, ....m ,xm,,^-'mI, =.f'4 [, [ _jj .b h. b ),. _f u,,t [ EI-A .u u,m m m _ m.'., m,.. m., mm, 4 u, N., % %4,,,
t 4 f-GE Nuclear Energy i i 23A4800 l i Kesiston 1 Class I September 1990 t i .t .. t 23A4800, Rev, I Supplemental Reload Licensing' Submittal for Pilgrim Nuclear Power Station Reload 7 Cycle 8 t f\\ 1 l 4 1 '/ / Approved: \\%tj Approved: ,1. S. h y, Enager E. ansen, Manager Fuel Li s Reload Nuclear Engineering 9 l ) 4
Pilgrim zwsm i Reload 7 Rev 1 i Important Notice Regarding g Contents of This Report l - Please read carefully 1 This report was prepared by General Electric Company (GE) solely for Boston Edison i Company (BECo) for BECo's use with the U. S. Nuclear Regulatory Commission (USNRC) for ~ amending BECo's operating license of the Pilgrim Nuclear Power Station. The information' contained in this report is believed by GE to be an accurate and true representation of the facts known, obtained or provided to GE at the time this report was prepared. 1 The only undertakings of GE respecting information in this document are contained in the contract between BECo and GE for fuel bundle fabrication and related services for Pilgrim Nuclear Power Station, and nothing contained in this document shall be construed as changing said contract. The use of this information except as defined by said contract, or for any purpose other than that for which it is intended,is not authorized; and with respect to any such unauthorized use, neither GE nor any of the contributors to this document makes any representation or warranty. (expressed or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor + do they assume any responsibility for liability or damage of any kind which may result from such use of such information. ? 1 t l-l t Pap 2 y 1
Pilgrim mom Reload 7 gn., Acknowledgment 4 The engineering and reload licensing analyses which form the technical basis of this Supple. mental Reload Licensing Submittal, were performed by E. O. Thacker 11 of the Fuel Engineering l Section. The Supplemental Reload Licensing Submittal was prepared by P. A' Lambert and [ verified by O. O. Jones of Regulatory and Analysis Services. r i l i l i 1 i .l 4 i i -1 I Page 3 E
i Pilgria m4ax Rg1 gad 7 Rev }' 1. Plant. unique items (1.0)* ' i Appendix A: Increased Core Flow Throughout the Cycle 2. Reload Fuel Bundles (1.0 and 2.0). ruelTyne escle I.maded Number Irradiated P8DRB282 5 24 P8DR"26511 6 60 P8DRB282 6 112 BP8DRB282 7 32 P8DRB282 7 160 New BP8DRB300 8 .122 4 Total 580 3. Reference Core leading Pattern (3.2.1) mwd /MT mwd /ST Nominal previous cycle core average exposure at end of cycle: - 18,362 16,658 Minimum previous cycle core avera from cold shutdown considerations:ge exposure at end of cycle 18,362 16,658 Assumed reload cycle core average exposure at end of cycle: 21,018 19,067 Core loading pattern: -Figure 1 1-y'
( ) refers to area of discussion in General Electric Standard Appilcation for Reactor Fuel, NEDE 24011 P A 9, September 1988; a letter "S" preceding the number refers to the U.S. Sup-plement, NEDE 24011 P A 9 US, September 1988.
i rege 4 b
i 1 i Pilgrim zww Heload 7 Rev 1 I i ) 4. CQculated Core Effective Multiplication and Control System Worth. No Volds, 46 20 C (3.2.4.1 and 3.2.4.2) 1 Beginning of Cycle, ' K,,,,, Uncontrolled 1.116
~~a Fully controlled 0.962 Strongest control rod out 0.990 R. Maximum inert:ase in cold core reactivity with exposure into cycle, AK 0.000' 5.~~
Standby Liquid Control System Shutdown Capability (3.2.4.3) Boron. Shutdown Margin (AK) (ppm) (20'C. Xenon Free) 675 0.040 6. Reload Unique GETAB AOO Analysis initial Condition Parameters (S.2.3) 3 Exposure: BOC8 to BOC8+ 8767 mwd /MT (BOC8+7953 mwd /ST) Fuel Penkinc Factors Bundle Power Bundle Flow-Initial Dul&D Leal Radal And R Factor (MWt) ' (1.000 lb/hr) MCPR BP/P8x8R 1.20 1.62 1.40 1.051 5.454 108.6 1.44 t Exposure: BOC8 + 8767 mwd /MT (BOC8 + 7953 mwd /ST) to EOC8 Fuel Peakinc Factors Bundle Power Bundle Flow = Initial DniSD Leal Riidial Axial R Factor (MWt) - (1.000 lb/hr) MCPR BP/P8x8R 1.20 1.59 1.40 ' t.051 - 5.355 109.2; 1.47 7. Selected Margin Improvement Options (S.5.1) 8' Recirculation pump trip: No J Rod withdrawallimiter: No l Thermal power monitor: No . Improved scram time: No 1 Exposure dependent limits: Yes Exposure points analyzed: 2 1-g: Page5-
5 Pilgrim mum i Reload 7 Rev 1 8. Operating Flexibility Options (S.5.2) Single. loop operation: Yes Load line limit: Yes Extended load line limit: Yes Increased core flow: Yes Flow point analyzed: 107.5 % Feedwater temperature reduction: No ( I ARTS Program: No Maximum extended operating domain: No l 9. Core. wide AOO Analysis Results (S.2.2) Methods used: GENESIS Nuclear with GEMINI ODYN and GEXL i ACPR i. Flux Q/A Event (EEBR) (% NBR) BP/P8x8R Figure 1 Exposure: BOC8 to BOC8+8767 mwd /MT (BOC8+7953 mwd /ST) Load rejection without 651 125-0.32 2 bypass i Feedwater controller 663 132 0.37 '4 failure 1 i Exposure: BOC8+8767 mwd /MT(BOC8+7953 mwd /SY) to EOC8 Load rejection without 730 126 0.35 4 bypass i Feedwater controller 718 134 0.40 5 failure 3 4 3 - Page 6
Pilgrim zwsm ' Reload 7 Rev 1 I
~~10. Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary J~~
~~(S.2.2.1.5) i Generic Bounding Analysis Results Rod Block ACPR Readine BP/P8x8R f 104 0.13 105 0.16 106 0.19 107 0.22 t 't 108 0.28 109 0.32 110 0.36 Setpoint selected: 110 r~~
~~11. Cycle MCPR Values (4.3.1 and S.2.2)~~
Safety limit: 1.04 Non.nressurization events: Exposure range: BOC8 to EOC8 BP/P8x8R - t Loss of feedwater heating 1.21' Fuelloading error 1.23' Rod withdrawal error 1,40-Pressurization events: Ontion A Ontion B f BP/P8x8R . BP/P8x8R .I Exposure range: BOC8 to BOC8+8767 mwd /MT(BOC8+7953 mwd /ST) 1 Load rejection without 1.42 bypass Feedwater controller 1.48 failure
~~The minimum MCPR value required by the ECCS analysis is 1.24.~~
" Option B is not available for this exposure. g ~ Page 7.. + n
t g Pilgrim zwax l Jteload 7 Rev 1 t f
~~11. Cycle MCPR Values (4.3.1 and S.2.2) (continued) l 5~~
~ Pressurization events: l Ontion A Ontion B BP/P8x8R BP/P8x8R s { Exposure range: BOC8+ 8767 mwd /MT (BOC8+ 7953 mwd /ST) to EOC8 load rejection without 1,45 1,40 .i bypass Feedwater controller 1.51-1,41 failure
~~12. Overpressurization Analysis Summary (S.3)~~
~~Pd Pv i Event (pgig) '(pg[g) Plant Resnonse MSIV closure (Oux scram) 1317 1333 Figure 6 1 i~~
~~13. Imading Error Results (S.2.2.3.7)~~
~~Variable water gap misoriented bundle analysis: Yes Event M~~
i Misoriented fuel bundle 0.19

14. Contrul Rod Drop Analysis Results (S.2.2.3.1)
Pilgrim Nuclear Power Station is a banked position withdrawal sequence plant; therefore, the control rod drop analysis is not required. NRC approval is documented in NEDE 24011 P A.US, September 1988. I i i i L j lt Page 8 '
+ Pilgrim zwm Reload 7 Rev 'l I i i
~~15. Stability Analysis Results (S.2A)~~
Pilgrim Nuclear Power Station is exempt from the current requirement to submit a l cycle. specific stability analysis as documented in the letter, C. O. Thomas (NRC) to H. C. Pfefietien (0E), Acceptancefor Referencing of Licerning Topical Report NEDE 24011 Rev. 6, Amendment 8, 'ThemsalH,giraulic StabilityAmendment to GESTAR H,' April 24,1985. Pilgrim Nuclear Power Station recognizes the issuance of NRC Bulletin No. 68 07, i Supplement 1, Power Oscillatioru in Boiling Water Reactors (BWRs), and will comply wl.h the recommendations contained therein. l
16. less of coolant Accident Results (S.2.5.2) i LOCA method used: SAFE /REFLOOD/ CHASTE See Loss of Coolant Accident Analysis Report for Pilgrim Nuclear Power Station, NEDO 21696, August 1977 (as amended).
i .j 0 I t I L 1 Page 9 - I
ein[d I MMMMMMM MMMMMMMMM MMMMMMMMMMM i
MEMMMMMMMMMMM
M M M M M M M M M M M M E

MMMMMMMMMMMMM
'::MMMMMMMMMMMMM ':M M M M M M M M M M M M E '::M M M M M M M M M M M M M l::M M M M M M M M M M M M M MMMMMMMMMMM N.N N UN 80E+DN.bNEbss.b.bD$$h B I 2EsE@e easeems s i MMMMMME IIIIIIIIIIIIII 1 3 5 7 9111315171921232527293133353739414345474951 FUEL TYPE. A = P8DRB282 0 P8DRB282 B'. P8DRB265H E = BP8DRB282 C = P8DRB282 F = BP8DRB300. Figure 1 Reference Core 14ading Patteni l-i Pas' 10.
Pilgrim Reinad 9 mg:n ) an., J hr kt t$l* L ME @ fhlM50
~~MUI!!V~~
_h m. f -. ,% k\\ [ / \\ / I N \\ u. y /z: .~. =r. Ild (Etact). I !f! gf j, /[' i, o v N [...& / W 1 E n d 5,,_, \\ i r ., s, vin cuc.cs3 tid (Ecoes) Figure 2 Plant Response to load Rejection without B (BOC8 to BOC8+8767 mwd /MT[BOC8+ypass7953 mwd /ST)) Page 11
Pilgrim - mam Reload 7 Rev 1 t i N !( 3 x:. \\ g,,,,je \\ f e-s \\ n .,...Il_ L.
~~n..~~
j tid (Mtht$) tid (Ktats) i I t f tt i ...y = i, / m - m yu e t E I... ,1%N 1 1 .i n... n.. tid tutocs) tid (ucacs) i Figure 3 Plant Response to Feedwater Controller Failure I. (BOC8 to BOC8+8767 mwd /MT [BOC8+7953 mwd /ST)). i Page 12
Pilgrim m. Reload 7 g,,. I dF j [ t \\ e --4, AN / sx 7 [ \\, l.:. .k. a. .r. 7. 3. ) ild (MCM't) ind (ECDct) < tpyrio-ey.w.s.ai> sp.g 5.; ni, b' E I UI f-y u. I& 'A j i () y l =. - ..n , p\\ n hy /.- t N/ h , W
-~
=i t J,. \\ f~ f l l t k ~s. i.e cucxsi i,4 iMe.esi t 8 Figure 4 Plant Response to Generator land Rejection without B pass 3 (BOC8+8767 mwd /MT [BOC8+7953 mwd /ST) to EOC8)- i L g Page 13 ^
1
~~!I 3~~
~~1 ,( 8 Pilgrim m. - Reload 7 ,,, 3 1-. hU r Ii 4 G,,,,~~
~~,n, w~~
" W,e -+ \\ s ^ 4 f ~ \\ i j[ L t-1 sia cucxsi t,,,,c 3 J_ j !f)hb5kh N!;ll$l' ' F i I l M ^, g 19 ...~ .,.g 4 9 l = i .:i 5: { s j-Li g. 3 a s. I i .i n.. a. I'd (MC6CS) tag gue,es3 i I (- j l + I ' Figure 5 Plant Response to Feedwater Contmiler Failure (BOC8+8767 mwd /MT [BOC8+7953 mwd /ST) to EOC8) g; Page 14
~~. l qr~~
8- ' Pilgrim . mesm f . Reload 7 Rev 1: 1 Kfh""a 1l!$!R!@U""~ r. E... 3..% / ~ I s x 3= c +. tid ($CCMS) - tid (ucees). @@lIi$f" d @ $ $ !!j!'i .l. t 2 1 ..s g lm a h,. . i... 1 o :. t 11 L1
==. .w., - taw (gemes) tid (u:ees) 1
~~n Figure 6 Plant Response to MSIV Closure, Flux Scram l~~
i: Page15 l
i 1 I ) Pilgrim - zwax Reload 7 Rev. 1 ' .a Appendix A - t ~ Increased Core FlowThroughout Cycle 8i l The analyses performed fo.; Cycle 8 included increased core flow throughout th'e cycle, l { There. are no concerns regarding reactor internals pressure drop of flow induced vibration as ' discussed in the increased core flow analysis document for the EOC 6 (NEDO 30242).- y The flow biased instrumentation for the rod block monitor should be signal clipped for a. f setpoint of 107%, since flow rates higher than rated would otherwise result in a ACPR higher than; reported for the rod withdrawal error.- I '. I I l 'i e s r i t' j f '( i, C I- 'I; '1 Y t 1 Page 16 l (Fina0 -
10CFR50.90 sosnwawsow Ngnm Nuclear Pour station. Rocky Hili Road' Piymouth Massachusett> 02360 s j Ralph G. Bird l semor vice President - Nuclear BECo 90-136 ~: November 8. 1990 l U.S. Nuclear Regulatory ~ Commission-l Document Control Desk-Hashington, Oc 20555 t License DPR Docket'50-293 ] PROPOSED 7ECHNICAL. SPECIFICATION CHANGE MINIMUM CRITICAL POWER RATIO. Boston Edison Company proposes the attached revisionsito Appendix A ofc f Operating License DPR-35 for'the Pilgrim Nuclear Power Station in accordance; with 10CFR50.90. The. proposed revision to the Technical Specifications upgrades the Safety Limit Minimum Critical Power Ratio'and revises the Operating Limit Minimum Critical Power Ratio. t These proposed revisions will extend the itse.of spectral shift'to maximize fuel utilization during the present fuel cycle. This change is not needed if our- -I proposed change of August 21, 1990 is approved first. ' ,, -) $f' w ,y.. rd , ( RAH /njm/4820 Commonwealth of Hassachusetts) County of Plymouth -) Then personally appeared before me, George H. Davis,-who bsing duly sworn,-did' state that he is-Vice President -_ Nuclear Administration of Boston Edison Company and that he.iseduly authorized to~ execute and file the submittal contained herein-in the name and-on behalf o.f Boston Edison Company and=that the statementi in said submittal-are true to,the best of his_ knowledge and belief. My commission expires:'O M ' C / 9 95~ b /# ev s
~~DATE~~
f NOTARYPUfIC Attachments: A. Description of! Proposed Changes B. Replacement Technical Specification Pages 1 C. Marked-up Technical Specification Pages. D. Supplemental Reload Licensing Submitted for Pilgrim Nuclear: Power Station reload 7 Cycle 8, 23A4800 Rev.= 1' I signed original and 37 copies .cc: See next page i >t
I J BOSTON! EDISON COMPANY-U.S._ Nuclear _Regul'atory Commission i Page 2' o cc:. Mr. R. Eaton, Project Manager-Division of Reactor Projects - 1/II: _ . a Office-of Nuclear Reactor Regulation: Mail Stop: 1401 U. S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockville Pike Rockville, MD-20852 l U. S. Nuclear Regulatory Commission f Region'I_ t. 475 A11endale' Road = King of Prussia, PA-19406-Senior NRC Resident Inspector Pilgrim Nuclear Power Station j i Mr. Robert M. Hallisey, Director: Radiation Control Program. Massachusetts Department.of Public Health-150 Tremont Street, 2nd' Floor Boston, MA 02111 B -l ? i I s l I
_._ _ _ - ~ - _ _ - _ _ _ - _ - _ - _ - _ - - _ _ - - -. .) .j i Attachment A' to BECol 90-1361 Descrintion of Pronosed ChADER 1 Pronosed Channes j Boston Edison Company proposes to upgrade the. Safety Limit Minimum Critical-1 Power Ratio (MCPR) and revise the Operating Limit Minimum Critical Power' Ratio. The revision to the Safety Limit Minimum Critical: Power Ratio reflects improved fuel designs in the core and the General Electric Boiling Hater Reactor Generic Reload Fuel Application NEDE-24011-P, i The revision to the Operating Limit Minimum Critical Power Ratio is being-proposed to be consistent with the upgraded Safety. Limit MCPR-and to reflect. spectral shift operation. Spectral shift operation results in, top' peaked flux. distribution patterns toward the end of the fuel cyclec A top peaked flux. distribution pattern increases the change in the Critical Power Ratio (ACPR) associated with the analyzed operational transients. The proposed revision maintains conservative operating limits. a The upgraded Safety Limit MCPR and>the: revised Operating: Limit MCPR ensure the. plant will be operated safely and will not. pose an undue risk to the. health andL safety of the public. These revisions will allow the plant to be' operated-more 1 efficiently. Basis for Chanae-The NRC approved the upgraded Safety Limit MCPR in.its' safety. evaluation for Amendment 14'to NEDE-240ll-P-A, " General Electric Standard < Application for r Reactor Fuel", dated December 27, 1987. Incorporation of the upgraded Safety. ( Limit MCPR is appropriate.for BHR's with D-lattice fuel assemblies provided: l'
1) the fuel has a beginning of life R-factor greater than or equal to.l.04 and consists of fuel types P8x8R, BP8x8R, GE8x8E or GE8x8EB; 2):the-fuel-is at least.2.80 weight percent U-235 bundle average enrichment; and 3) the' lower; enrichment-bundles in the core have operated for at least 2 cycles.'
The fuel assemblies presently in the core of the Pilgrim Nuclear Power. Station satisfy the conditions evaluated in.NEDE-240ll-P-A.. These conditions will also i be satisfied for' future reloads. M L The Supplemental Reload Licensing Submittal for Pilgrim Nuclear. Power Station. l Reload 7 Cycle 8 submitted by BEco letter 87.081-' dated May 22,L1987 has been revised to reflect operation using a' spectral shift fuel management strategy.. ' The revised analysis also reflects an assumption that the' Turbine: Bypass Valves do not open in the analysis of the Feedwater Controller Failure'with maximum demand. This assumption was made because precise measurements of.the opening; a time are not.available. The-assumption is conservative and ituresults.in a r more severe transient analysis and a more conservative.0perating Limit MCPR. This limit is incorporated in the pronosed Technical Specification. 'The' L proposed Operating Limit MCPR will er the MCPR'does not. decrease below the proposed Safety Limit MCPR at'any time-curing any abnormalJoperating transient as defined in the FSAR. l l 1 .m. e
1 Determination of No Sianificant Hazards C6nsiderations' The code of Federal. Regulations (10CFR50.91) requires licensees requesting an l l amendment to provide an analysis, using the standards'in 10CFR50.92, that I determines whether a significant hazard consideration' exists. The following 1 analysis is provided in accordance with 10CFR50.91'and'.10CFR50.92 for the-l proposed amendment to Pilgrim's Minimum Critical Power Ratio. L 1. Upgraded Safety' Limit MCPR A. The proposed change-does not involve a significant increase in' the-I probability or consequences of an accident previously evaluated because the l proposed change to the Safety Limit MCPR does not change plant equipment, operating procedures, or governing design criteria'used to protect the plant against the initiation of any analyzed accident or used to mitigate the consequences of any analyzed accident. B. The proposed change does not' create the possibility of a new or different kind of accident from any previously analyzed because the proposed change does not change plant equipment, operating-procedures, or governing design criteria and the change to.the Safety Limit MCPR provides the same level _ of protection as the existing Safety Limit MCPR against fuel cladding failure-during an abnormal operational transient. The proposed Safety Limit MCPR therefore provides equal assurance against a release.of_ radioactive material in excess of 10 CFR 20 limits _during-abnormal operational transients and a new event sequence leading.to an accident is"not created. C. The following design requirement ensures an adequate ~ safety' margin'is-maintained: 1 Abnormaloperationaltransientscausedtiyasingleoperatorerroror equipment malfunction shall be limited such that, considering uncertainties in manufacturing and monitoring the core operating state, more than 99.91. of the fuel rods would be expected to avoid boiling transition. The proposed change does not involve a significant reduction in the margin l ( of safety because this design requirement, which governs fuel cladding I integrity and maintains the defense-in-depth philosophy; has not changed. .2. Revised Operating Limited MCPR ( A. The proposed change does not involve a significant Lincrease in the probability or consequences of an accident previously evaluated because the proposed change to the Operating Limit MCPR does not change plant equipment, operating procedures, or-governing design criteria used to protect.the plant against the initiation of any analyzed' accident or used-1 to mitigate the consequences of any analyzed accident. j p d l
t >l i'I B. 'The proposed change does not create the possibility of a'new or different kind of accident from any, accident previously evaluated because.the: proposed change does not change plant equipment, operating' procedures,.or 1! governing design criteria and the changes to Operating Limit MCPRs provide-T
~~the same level of:assuranceLthat the Safety Limit MCPR will not;be: exceeded 9~~
during an' abnormal. operational transient. thereby assuring a release of radioactive material ~ in excess of 10 CFR 20 limits;will not occur during ) abnormal operational. transients and~a new event ~ sequence 11eading;to an, accident has not~been created, j C. The proposed change does,not involve a significant! reduction in the margin; of safety because the conservative Operating Limit MCPR' ensures the most ' limiting transient will not violate the Safety Limit:MCPR. Reauested Schedule The. requested change is not needed: Until the present fuel burnup reache's 7200 U MHD/T, about January 31. 1991.- Additionally -approval of~this request will not be needed if our proposed change off August 21,.1990 is ~ approvedifirst.; v' y 'l l 4 4 i ~ l o s i 'h .n I> h' n 1 I 3 'l
-~ Attachment B'to BECo 90-136 List of Effective _.Pages l i Revised Pages.. 6' 12 7 13 205B-2 ( D 4 i .'t s 9 ' i .4-l l. d l l ' k l l-1 1 1 I . f 9,l '( g; l-l l l> e w w
'[ 1.1 SAFETY LIMIT-2~1 LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY- ~2.1=-FUEL CLADDING-INTEGRITY ADD 1icabilitv: ADD 11 cabili tv:. Applies to the' interrelated - Applies: to. trip settings tbf:the variables associated with fuel-instruments and devices-which~are thermal' behavior, provided to prevent the reactori system safety limits from being. Obiettive: exceeded. ~ To establish-limits below which Obiettive: the integrity of the fuel clad-ding is preserved. To define.the'1evel of the' process = variables-at which automatic < - Soecificatich: protectiveactionisinitiatedtof + ~ prevent the fuel ~claddingL 1 A. . Reactor Pressure >800 osia and integrity safety limits from being-Core Flow >10% of Rated exceeded.- The existence'of a minimum Soecification: critical power ratio (MC?R) less-1 ( than 1.04 shall constitute A. 1 Neutron Flux' Scram: 1 violation'of the fuel cladding: integrity safety limit. A MCPR l' of 1.04 is hereinafter referred The limiting' safety system trip r to as the. Safety Limit'MCPR. ' settings.shall be-as specified below: B. Core Thermal Power Limit (Reactor f Pressure:1800 psia.and/or Core -1. Neutron ' Flux Trio Settinas 1 Flow 110%) a. APRM' Flux Scram Trio Hhen the reactor pressure is Sgt.p no-(Run Mode)- I 800 psia or core flow is less than or equal to'10% of rated, When th'e Mode' Switch is in the steady state; core thermal. the-RUN position,r the APRM power shall not_ exceed 25% of flux, scram trip' setting-design thermal power. shall be: C. .Eqwer Transient S AL.58H + 62%'2'1000 The safety limit shall be assumed Where:- H to be' exceeded \\when scram is known to:have been accomplished S'- Setting in percent of by a'means;other than the-rated thermal power l expected scram' signal unless (1998 MHt)~ analyses demonstrate that.the' fuel cladding ' integrity safety -H - Percent ofidrive flow = limits defined in Specifications. to: produce a? rated core 1.lA and:1.lB were not exceeded-flow of.691M lb/hr.. during the actual transient. Revision L6 n Amendment No. 72-1 I; r
4 BASES:. The statistical. analysis used to determine the.HCPR safety limit is. based on a~ model-of the BHR core which simulates:the process computer function. Thereactorcoreselectedfor.theseanalyseswasa'largel704c assembly, 251 inch reloadicore. Results1from the large: reload' core-analysis apply for all-operating reactors for allereload cycles,_ l4 including equilibrium cycles.. Random Monte Carlo selections of all. operating parameters = based on the uncertainty ranges of. manufacturing tolerances, uncertainties in measurement of core operating parameters; calculational uncertainties, and statisticaliuncertainty associated'with-the critical power correlations-are imposed upon the analytical-representation of the core,and the resulting bundle' critical' power ratios. Details of thisLstatistical analysis are presented in Reference 2. B. Core Thermal Power' Limit (Reactor Pressure < 800 nsia or Core Flow-A 4 < 10% of Rated) ~ 1 The use of.th'e GEXL correlation is;not valid.for the critical power' calculations at pressures below 800_ psig or core flows-~1ess than:-10% of rated. _Therefore, the fuel cladding integrity safety limit is: established by other means. This isudone by establishing al limiting _ J condition of core. thermal power operation;with the following' basis. .d Since the pressure drop in the bypass region is essentially all elevation head which is 4.56 psi the core pressure drop.at low: power and' 4 all flows will, lways be greater than 4.56 psi. ; Analyses show:that with' 1 a_ flow of 28x10 l's/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a..value of 3 5' psi ~. 'Thus,-the 3 .] bundle flow with a 4.56 psi driving' head.will.be greater than 28x10. 3 lbs/hr irrespective of total core flow and independent:of bundle power ] for the range of bundle powers of concern. Full scale' ATLAS test data, n taken.at pressures from 14.7 psia to 8000 psia ~ indicate:that the fuel. assembly critical power,at this flow is approximately 3.35 MHt.' With .t the design peaking factors the 3;35 MWt; bundle power corresponds,to a: L + core thermal. power of more than 50%.j Therefore a core-thermal power-limit of 25% for: reactor pressures: below 800 psia,. or core flow'less l j than 10% is conservative. l l' n 1 L i t L 1 l r 4 Amendment No. 42-12' n
I BASES: l C. Power Transient Plant safety analyses have shownLthat the scrams caused by exceeding any safety setting will assure that the Safety Limit of Specification; 1.lA or 1.lB will not be exceeded.- Scram: times are checked' .l periodically to assure the insertion ~ times.are adequate. The thermal) power transient resulting when a scram-is accomplished other than by the expected scram signal-(e.g.. scram from' neutron flux following! closures of the main turbine stop valves) does not'necessarily-cause, fuel damage. However, for this specification a Safety Limit violation: will be assumed when a scram is only accomplished by.means of:a backup' n feature of the plant design.. The' concept of not approaching a Safetyi -l Limit provided. scram signals are operable.is supported by the extensive plant safety analysis.. ~ The computer provided with Pilgrim Unit'I has aisequence.annunciatio'n' program which willaindicate the, sequence.in which events'suchi as: scram,c q APRM trip initiation, pressure scram initiation, etc'.,' occur. This. program also indicates when.the scram setpoint-is cleared. This will: provide information on how'long a' scram'. condition exists and'thus-1 provide some measure of the energy added during a' transient.- .i D. Reactor Water level (Shutdown ConditionP During periods when the reactor is shutdown consideration must also be l given to water level requirements..due to!tne. effect ofsdecay heat. gIf-reactor water level should drop below the; top of theLactive fuel during a this time, the' ability'to cool the' core is reduced. This reduction in .i core cooling captbility could lead = to' elevated cladding' temperatures and clad perforation. The core can be cooled sufficiently should.the1 water level be reduced to two-thirds < the, core height, c Establishment.of the safety-limit at 12 inches above the top lofgthe fuel;provides-adequate margin. This level will~be continuously monitored. References
~~1. General Electric Thermal: Analysis Basis!(GETAB):' Data, Correlation.and' j~~
~~L Design Application, General Electric'Co. BWR SystemsLDepartment; November 1973 (NEDO-'10958).~~
~~2. General Electric Boiling' Hater Reactor Generic Reload fuel Application,-~~
NEDE-240ll-A. i l Revision ' Amendment No. 42 13L j I '.l. ' -. = e' e e ^ - +
V ] i TABLE 3.'ll-1. OPERATING LIMIT MCPR VALUES j -l A. MCPROperatingLimit~fromBeginningofCycle(BOC);to[BOC-+:7,953MHD/ST.. P8x8R/BP8x8R! For all values of x-1.48-B. MCPR Operating Limit from-BOC + 7,953 MHD/ST to End:of Cycle, t - P8x8R/BP8x8R g t 110 '1 41; 0.0 <'t 1 0.1-1.42 0.1 < t 1 0.2 - 1.43 0.2 <.t 1-0.3
~~1.44 0.3 < t_110.4 1.45~~
~~-i 0.4 < t'I'O.5 1.46 -0.5'< x 1 0.6'- '1.47: t 0.6 < t 1 0.7~~
~~12. 4 8~~
..i 0.7 < t 1 0.8 1U9 Ji 0.8 < t 5 0.9 1.50! 0.9 < t s>1.0 1.51-l l 1 ^1 j Revision- ' Amendment No. 108= 2058-2 1
i 4 lI . ATTACH ENT C to 'BECd 90- 136l- ~ Marked Vo~ Technical Soecifications i t - f I f ,- ( f ? i s h i s W 1 s ') i k L o i .f f -i I - 4, ; -
'2.1 : LIMITING SAFETY SYSTEM SETTING 1.1 SATETY LIMIT' 1.1 ' FUEL CLADDING INTEGRITY-12.I'FUELCLADDINGINTEGRITY Applicability: ( Applicability:' Applies _to the interrelated. Applies to trip settings of:the-instruments and devices.which are- ,U variables associated with fuel thermal behavior, provided-to prevent the reactor s -f system safety _ limits from_being 'i exceeded. Objective: i; Objective: i 1 To establish limits below which, To(definethelevellof-thepro- , cess variables at which automatic the_. integrity of the' fuel clad - ding is preserved. l . protective; action,istinitiated.to prevent t.he fuel. cladding inte- _7 grity safety limits from being_ exceeded.: Specification:- Specification: -i A. Reactor Pressure >800 psia and' 'A. Neutron Flux Seram: ~ Coce Flow >10% of Rated s The existence of a minimum-. The! limiting safety system trip-i critical power ratio (MCPR)' settings shall_be;as specified-shall'consti-- below: j, g less than tute viola, tion of the-fuel- . cladding integrity ty. ~ 1. Neutron Flux Trip Settings limit. A MCPR of is here-jg inafter referred to as the ' a. APRM Flux Scram Trip l Settina (Ran Mode) Safety Limit MCPR. I B. . Core Thermal Power Limit (Reac-When'the Mode Switch ~is tor Pressure 5800 psia and/or-in the RUN position, the'APRM flux-scram-l- Core Flow $10%) j ' trip' setting shall be: When the reactor pressure is i S S.58W + 62% 2' loop. 1 800 psia or core flow is less than or equal:to 10% of rated, _t l l the steady state core thermal i Where: l power shall not exceed 25% oft S-= ~ Setting in percent L design thermal power. Lof. rated thermal power (1998;MWt)- l I C. Power Transient t p L The safety limit shall beJas- . W= Percent of drive flow to produce a sumed to be exceeded when scram rated' core flow of' -l l L ~is.Known to have been accomplished..- L by a means other than the expected ' '69 M.Ib/hr. L scram signal unless analyses demon- ' strate that the fuel cladding I, 1 integrity safety limits defined.in. Specifications.l.lA and 1.1B were l l' not exceeded during the actual transient. 6 Amendment No / u n ~-
.:3 ..]' 7 .h y)32 W "g' O ; r e s .d A.3,,w i Lw U. L. Lie Lie.1 dil . de w.~eI L.1;s.1=s / 11 nd ;; Tebi; 5--1, M f: n n 1, th: r9 r:1 - ef de x:: j -Petametere-14*t;d i; 7;ble " 1. Rif e.-ee 0,.ud G.,6.1.6.ivu 7 z s // j n; L17 pear dinriktia eh. 2 Tir.e ' 1.od ' ' _ vi a_. a \\ . &_w_ ..n=6._. .r ^d*' R f e;ter distributione-thet n;
~~Tut tr *'_"..St S*d "*d *"'. ".~~
.i ~ \\ >a .w...._.wAAmu..Aas p.4. iJ .s.AasA h n w r n. 444. arass hw& _. i w asaw u de wwww bv 1 -distelbweden; ;h;--. ;;; I;h;; ;;n ;h h;' ';;; ;f th f.;l .y. O h:i: fer de reert ' tic; de dr =:; p=r:t=: =; ;;i;;; I i; 3 0 203'O
d 0.; h ;is in O.; ;;eesT=4'sy 1; ;,1.e,

ll;.
eece*4*t4en-ie-given i.; D1M5d". O ; p s ;; di;trik ': ~l is h;;d-on ;,,7,.i;;l 'M n:M ay ;;;; i; -4.i;' 2: ::d';;t n. Wee ali;;;;ily chee;; I; pieda;; ; eks;d p; s dinriL.i;; LJing - th; gr;;e+et "a ;f an;tliee-e; :.he ti.Len pe.u 1.e.le. The
~~di:::iht10: 1: Pilgri; l'ah= rs;; Ot;t1;; '.';it i duf;;~~
eey-fwel-eyel; =ald en L. e; e.4ere L.L. Ji.L.16Livv.. L. ii 3 Core Ther=al Power Limit (Reseter Pressure < B00 osit' or Core Flov ' < 10% of Kated) i The use of the gen. correlation is not valid for the' eritical power calculations at pressures belov 800 psig or-core-flows less than 10% of rated.. Therefore,gthe fuel claddf ng integrity saf ety 14' t is et 4 11shed by other means. This-is done.by-establishing a 11. ting condition.of core thermal power operation. with the following basis. Since the pressure drop in the bypass region is essentially all elevation head which is 4.56' psi the core pressure drop at lov-2 Analyses show that with a flow of 28x10ger than 4.56 pai. power and all flows vill always be gran '1bs/hr bundle flow,, bundle pressure drop is nearly independent of bundle power-and driving head vill be ' greater than 28x10ge flow with a 4.56 ps1 has a value of 3.5 psi. Thus,'the bund l i lbs/hr irrespective-of total core flow and independent of. bundle power. for. the range of bundle powers of concern. Full scals ATLAS test data takan 1 ~ at pressures from 14.7 psia to 800 psia indicate that the~ fuel asse=bly critical power at this flow'is spproximately 3.35 Wt. With the design peaking factors,the 3~.35 W: bundle power cor-- 3 responds to a core thermal power of, more than 50%. Therefore a core thermal power li=1t of 25% for reactor pressures.below-800 psia, or core flev less than 10: is conservative. C \\ i e W -,er .e -a w ,e ,av-,,
I n s e rt A' The statistical analysis used to determine the MCPR safety limit is-based on a model of the BWR core which simulates the= process computer function..The reactor core selected for these analyses was,a large 764 assembly, 251 inch-Results from the large reload core analysis apply for all-reload core. operating reactors for all reload cycles,. including equilibrium cycles. Random Monte Carlo selections of all operating parameters based on' the. uncertainty ranges of manufacturing tolerances, uncertainties-in measurement of core operating parameters, calculational uncertainties, and statistical-uncertainty associated with the critical power correlations are imposed.upon' I the analytical representation of the core and the-resulting bundle critical power ratios. Details of this statistical analysis are presented in j Reference 2. 4 } I
...t::=- -t " -- - w e .1 7 o a i i 1 1 1 C. Power Transient f Plant saf sty analyses have shown that the scrams caused by ex-ceeding any saf ety setting viu assure that the Safety Limit of' Specification 1.1A or 1.13 vill not be exceeded. Scram times-are checked periodically to assure the insertion times ars j adequata. The thermal power transient resulting when a scram j l 1s accomplished other than by the expected scram signal- (e.g.,, scram from neutron flux fo noving c.losures of the main'turbins l stop valves) does not necessarily cause ' fuel damass.: Bovaver, ' l f or this specification a Saf ety Limit violation vin' ba assumed - 1 vban a 'scraz is only accomplished by maans of a' backup fastura: of the plant design. ~. The concept of not approaching a. Saf ety L1=it provided scrs= signals are operable' is supported by.the s' extensive plant safety analysis. r The computer provided with Pilgrim Unit 1 has a sequence : ~ annunciation program which viu indicate the-sequence in which events such as scram, ~ APRM trip' initiation,' pressure scram initiation, etc.' occur. This program also indicates when the scram seepoin: isicleared. This viu provide inf onnation' on hov icng a scrac condition exists:end thus provide soma'maasure of the anergy added during a transiant. 1 D. Reactor Vater Level (Shutdown Condition) ' l During periods when the reactor is shutdown, consideration must also be given to vater level requirements due to the affect: If reactor water level'should drop below the ~ of decay beat. top of the active fuel during this tima,. the ability to cool-g the core is reduced. This-reduction in core. cooling capability; could Isad to elevated cladding-temperatures and clad.parforation. Q The core can be cooled sufficiently? should the vatar level be ~ L reduced to tve-thirds the cera height.- Establishment of the = saf ery - limit at 12 inches. above the top of the fuel' provides' adequate. This level win be continuously monitored. margin. References 1 i General Electric Thermal Analysis Basis (GrIAB):. Data, 1. Corralacion and Design Application. Ganaral Electric Co. - November 1973 (NEDC>-10955).- BWR Systems Department. m r l '." T ;;;;; 0._pt:.r ?:rf;r m: O rir-*irn /20" 207. C rr ni-41::trii C : piny 3 4 &yet--- 0:p r r t, 220 19 % i p C O O 200'0). / General Electric Boiling Water Reactor -Generic Raioad Tuel Application, NEDI-24011-P. -A 13- + n ,'y.
l ' TABLE.3.11-l' OPERATING LIMIT MCPR VALUES 7 95 1 A. MCPR Operating Limit 'from Beginning-of Cycle (BOC) to BOC 4 WMHD/ST. 'P8xBR/BP8xBR-l i 7,9 @j,48 For all values of 1 f B. MCPR Operating Limit from BOC + MHD/ST to End.of Cycle. / t Pax 8R/BPBxBR T !. 0 A.Yl l / .h 0.0 <t1 0.1 h40- /, M2 f 0.1 <tS 0.2 }.Y3 0.2 <1 1 0.3 /* YY h /,4 6 / O.3 < t1 0.4 h / Mb: ( 0.4 < t1 0.5 0.5 < 11 0.6 / $?: O.6 < 11 0.7 /ed p 0.7 <t1 0.8 -H M ),}f V y 0.8 <t1 0.9 4:44- /.6d f 0.9 < 11 1.0 i /, 6'/. / j 4 i l L i l Revisicnh 205B-2 W (kr.
m - w i' WQ s.iQ3V WY Q i ', - ~.~ - w~}_~i -, Rh t p, e ~ n' ' @M e allQ. n-a f n g) $a.1&c}w &.%~ QMO W M J Q,W W.hyL w M m m wwm mm ~ n~ m. ~' T-QQ [ *Mi Ub ' ' ~ 4 y;f J r_ h y q 4Wu wwmMWwe &[#+ b [g:a.badh M;e nww,hh,N c .A:hh on nAw~uhkn.x,hkhh h h hh hkN k hh nn an wmww wk e
~~n. w m m., e, + m mo n, u,~~
~~x pgenna - 4u ma s,: a pp. .mn as c n nn m v~~
~~n. ; ;y c~~
~~w,g wwny; g y qws.pu%mmy upc yyy u e gg yp. m sy~ e 4 ~ s~~
~~9. o g,-r gm,.~~
~~m wm:y gm.c ug n a wdgm;;k w vamn w p na yg w;pqq, y p;qmhww w;; pnp Qp e y .. m+: sww?w w m - + -_s n ~a -,mu a Q. m t c ' ~ m.s. ;u-c, y 3g+M y c T.wy g} gg%... a- =g 5~~
~~t h w r e- ~ -- 4 g %g,,am. g~~
~~'7g, W W Q..n D. r.Y -; '--?. 2~ V tl~~
~~e. %.~~
~~m ,s. ^ > '1 -- k +-uq o~~
~~t.u. -. - v.O-v v.~~
~~4 s u 4:.. y s .~ m y g@MEm&.rm~~
~~n n;~~
@yy.j,M ATTACHMENTlDitonBEco190-136 M i9 W 3 Ri w / j ~ G+ gM C 4 .s y y. mv., L : , % D & ; Wa;JM M y% % Q..WW.%9 WPMb Supplemental Reload Licensing! Submitted MorfP11gr.tajh clear! Power E g W M & A p Q r1. y % '3 W
~~Station reloady Cycle;8N23A4800lReM1klg ? M9%f@#@$WP@$5MM e m,o, m+ s w v,n ; A W g g,_ g~~
~me p%u w .b.,#n~p hn a n nma - = .w .~ n 3 py. m. g a_qy a. n.s. p. y u y, u --. ~, w w.. 4(;Q'M,Qn ; j }i'. m v ,, u. s -.m mn . ~ - , h,m'f J".d_i^ 1' I-(OM. $ I.Y.s ? d.Y.:41M 1. N P b- \\ l'~ ~~ Mh Ed E s td o f ' #g ' a 1 _t 2 a e s ey ; m. v4 ~- s' n': W 7 g - f
' ! 6 (,.h ii w*
~~$a < J w i~~
~~r n~~
~~4 cu 1 M pfk yg s ,_g_ %f-~~
~~rm igm.~~
~~3 si F si~~
~~h ppg pe s _ gy,..e e Qu n w,-D N tP Ns &, M&.;L :j f:_~~
~~gt 7 N 4 e u 47 p'f n y%gg;p.,v g =c-4 i. a1 t < :e.7 a r . - - -m n.7., u, ; nn s e MWM - V u,"- %' ' '4F'~~
p.
~~J ',P 2'-' ,og L t g ^^ Q; Q yW jQx ygy~~
~~D r~~
~~y ^ ~_ M H a f.3^p[i tNuW. . i, g)-q H. I t q- )h 3 ,.x c. t z V_ - t45 pf _ ; } :;~~
~~n.,Lg t W y-l N~~
~~- p p. J.:p 4 Wg ':g e 3 N...N i R3;* 2 4 y g/ s'/ t.~~
~~+ tt f 7 4; & &~~
~~t { W y-. j' ; 2., gs,:,e ?, r6 1. 'L. ) n, WqO' ?'- na~~
~~sye g l, o~~
~~\\ ^'w 'w 2 3 6 o4 v@_pv Aww@-! 0~~
~~a. A 31u p-rp neW,,y r~~
t:1 m-i a' d - n p,p'o wego,,,W.. s I p, ~ - e wm p 9 - , %. a. ~ pm., s g :n a 3 frp%.. e, tqq:w%.gWW WA ace.y e < u,p ; g< N;g.r 3 4 i w m+: wgp ;m[c.m"1 < m+ w e m q w o m q. n y y ;,.. m~m r3* um . ~3n nw .ss, m w n:,> t w.W"y' W)a% f %a.. m,% nk- > u';m:@mwe gw n p s W;/%gqNg esN4pqv n /e.;m oi , : g; aww', y ,,.M gm aw g,' u c. ,
1%
g N y! g g. g g g g g<@n g,g %g% w, i 4 gp p & l 4 = 07: 3 v ,3 dQ![4.. upg#m@p-;ap@%@;Y!W%ggg%e, s M dM a .p' M h 1 c c m y p p y m y/T M h W 2 h ji y MuI-r Q i 4 =* .W s y 'w%,W GR- :4. y,gy g6 4 !@ .t.,3.m%o.-qgc a < m aye w%wp AmM c ,p c w@7 nm ' ~ -,.. x.,s w g q nsqAW . w_ a em p g g m.m wg y y - 9 g y Ny: 3 B:. w s.mgD - A,. g 2 y Mju % JMy ,.j u t -shm oyo. , r. 4, '1 w,. n: b 7 4.m,3 m- ,.v s yp w m w} +% {p} p --. ; r n', j ; i
~~n. 1 '_d J rm w^f 7:g gi a 19 y ?,g* ?~~
~~y+ c.u gf. , j- ..L r ,9.? Q u<i+ MPse O & y%~~
G.x, w, w g_gVwLn yt;;m n t > s - Mc e

Wvu
> - +.;d y 1 v% _. s. ,8-m n 7 > s q3 w s.s cgn s J; q w-: =, p 7.> wwgu m: r <:S RJ;o Wc q ~-% ,e .M k %w sp 4@m _ w .wm r 1 d M (w yh 9 r; app 3,yg xg, Mp@)m:p ;w&xnwawkw s7 i Ve WJ p m 6, Ti %.g, : p M. nw^ . g= ;, =,<,., - 4 hk px 47w nw, g gg-gn ygxQpy s s ; y-- [ 7 y f;gwamyggy, gg-
~~2 va~~
.r gMM)TTQ/Q,-Qngu[d pw pg ymg.f g.. 7 V..- fH r g' 'g < 4. gy, L,pg%p m>mggmgdky Q Q 2,gg$y%g.#lLfp"enelg$nw 4 -W " ilg g p' m 4 a gV i M o! w!'r t g s, W 9 r, m w t 6 G.a -d e ; 6QM Wh;M@%yG W s i W N. r_J-d
~~i c.~~
~~.w: i i ym @a i . m y g. ; M~~
~~l' \\~~
~~~O m#ygge m~~
~~9. h :', -s4mummgg-~~
~~.p n t e ag , wg +,. Jg ce y,~~
~~k +;v !y4.i ; je %w-W'r-M>i s,~~
M 'M W q"m u ; 6-k w,e, m..- 5 A"- u t %, p,J . - s l. v.p@ 3-v t >,i_ 1 s &a1-P Cf33 r m p kgv3 J ft. .is-4 J s? J a t% .,e?y - a = %.qtm:w v gw -yh .-w t hy8-y 1 i M q. u (&~ s 3 p% g 21 a r - i s / h w g(ge y Y ]g( y { w / g b %m e 1 p i Isc 2 g e..b < e M y Q :[% m Wy s m+' q. M pa w w%Uyj$, % -[f ad +4.p#'E^ Q % L s w.p:M m i o' 4 \\ yy -; c % . W. 'y ,m" M.m.%y y % n e
~~w. :s h,t ' h, '~~
~~6m q%, ,3 p. + s d; 6~~
~~s. :. 1 t~~
~u g y j 6d'0 4/ 6[W 3 Y ['N ' N . MlN ' ^N ' S-c k _ m pwmqh.M i @%; e.M[ 'M-3 g$$ff, - 6,.., ; T ' b, ')i 1 9 $ /M.i 5 Pp y.) 4 y t 1 t w m :w [ 4 i;(# / "p a d a[pwo% ; : w%. M r.m. m i r. : u t 4 shQ W x. p ,_ % y are sam 4 s n M m+.w w w -@M.-N h + 'c.d I h4 P'IE O M,;-QNh uW w h %n$Nm-a pQ.hk' fU' i d['.. 1 ., M. M $,1h4W' -}% d.; 7 -s g'I; W@i 5 Q q -41 t- ,.4-1- g-nQ< y a W j M g w MnN/m adgeM e g4jMu > u @W3 m': eg -- m 3" E n @rW @W C A "d-[W,J E"f W h (' J V)g N p u mg e 49
~~J;a~~
~~pg.~~
~~6_ Ay .b. o .n a s um 4* Mppg m ;s h'mw>c.4;w m,O.;% '3 3-a (WRM~~
4W
~~%1 W{ P L ch +- 4 ,e ,ae N^ ; t Wg.. r% ".alyi, - tej nj 9 7. Gm -i d M.~~
~~.M w~~
~~6 ?~~
~~. W' r~~
~~s. t c y e, 0 N?r 3gn n e m,y%w s m. wt i~~
f
~~.x, g&g %..,," %a'4 y v D..(. m t-i cL J p9 Jnp %y+a e. m,ne J., t. 1 - T/M'm, 4. %.. y _: u, n.4,. ; w u -. m, c (' ta1 A 'i .gh g.u q 3-2~~
~~W.. t ut l% t#'iM'~~
~~~ -, + ,a lh, 40 +- V Pl ' ' ' Vm ny { &(.f*l$llf@.; g g*/#, h n J M.; 3;i. v, ' }l. l *'^ ) ~ e kns ' "b N L$h. ^ f. %~~
~~5 i~~
~~f~ dp, h Q ,) ' ll, wpfl'd; l i t 4 1 W p riwl=- &.Anp %y'Wwy m, mw%y s 9 T. .~ 4 e -y g o -Mf N; u, -.; t y-h-'a.'- J t x,... c q( } {J m40-.] 2, g 3p'W % 4,; _ f[?.~~
~~- If~~
~~/ k ' ;a 4 -Q p< +3 3, " h. (j.-4]; s f r e:pf;,%p%; p%yf,;'m.MW ! . jp 1 p 1' f7 t y s,;' 3 j, ' 4' i pf 0 n ,g_g ),~~
~~3. s~~
~~--t m4 z.M., -... f, [m g? 4 q.y ,q yM.p&: ,c u,~~
~~. k. ';Q,'; y'~~
~~'* 4,.~~
m.. ^

- ?,,. a.a u n.

n.
i . ;, 4 7. q, ; b .s 't <l e .s a k g V.s,)_ ~ ( Q:. m g,m i[ j_ tf 4 1 's .. c ['A [g ft a = k.I,. m. 7 4 Y, ,.s m s . r:v,f.c]ht k, rh. .y , b E h .g;j I$ 4 I/ .g j 3 .., m M., _.- k ' 4 1; f,t: w 1 .J',. u$.. M.e kob=o, md !pg g y 5 j gql . fd i t e'cLN . nn ;[:p v li j m 3.a xWQ Q.y.j'E t yi v, 4++ o h:i., 3 m wmc A.- 1 , s., + - &. r.. :p.s m < -m: y :, mp wvo msg r; mnma,t v a t J :'; h./w k,*y! 3 s,e r p - ! \\; b ed
~~p. Mt'.F. m' w +,>~~
~~Q? ,s ;., 1 ~:. p. 8 ..., 4, a u s -V s; v .2 8 ..c,< I ;-- ,1 < m. %e.. + *, r d w e n,n I,,.,..t, Y' -q. f f r r p n g' ' 5' {" g V'-J%' g is**L 5(..4F p P-r"r -t l'.'c ?<..,-h[ #..~~
~~4t e~~
~~y , ' >. nj., t r 'n [q.,' h.E f 9 -, e y y, _", j., 'j.. @t.;, p-( 3v'M.9w, %}[y/ E n' [',.P 1 s ' h 3 % -, f. ( '. {. nI 'g~~
~~p" pg-d~~
~~"'.r 'm y ' U _ ~i : t .F <r .o 4 g c vfT :- '?a :: ' i g~~
~~alo di-y z. 'a.g~~
~~. ;, 1 y c s h.w-',.-it h A~~
~~3. N.c h q,.4 u~~
~~y y ,my: ,t-4 t 3 q- -.,3.m. m, g sg~~
~~7..;., q.,~~
~~m eu~~
~~q;< 41c r m m n~~
~~s ag g - v 4m. N . 3 - - t m g' :3.ct!. v1 %...j r3- < j i~~
~~1..~~
~~u 3,,.k,y r. n : < t 3su .a.q. ns u + .va j .r~~
~~hf 2 Ii. ;b ; Q[~~
.:.W, I 9,-[w. k : y ;[r c hJ[m. r, , fi'- ..-k [ n.- m u ia ,,. h f k I,yc--E F.*'h^ g r&' d ' b< i a _v. N 12 :j +r J%yhJ,i.s s a f4j ): h ~, ' M g w:q n n'oa c y :;s K-,n,Iy t., y qw =Q. ya, m.. r c-3ma1 e= >. h x n s g_v ~ ' M.. y y n '~3 :,3 a6 x i 4 5 . w.m y w,G u m m -:, M y' I :jJ',= [; 0 ?'1 i r 1, M.-,_3, i+3 d I. ; Jyy g, q q,.qu .1 \\ ?. O fpj gi..]., .in ieA iM: eg. M f ) 3 r. 37 4 7y
~~. s un !.~~
~~i 'r q t 3 c d~~~
~~6 e.~~
~~7 i s y.. e - 1 sw b i \\ i 2[ m.s p ( n~~
~~- mans b I l~~
~~}( =-g' S[ n m 'v 1 n.w.~( m _w,, u. ', j y~~
M.

A n -
.e . e. e m n e,. g,1 fj.m<;:. g,u.&g ; w,, %~;, %m+Wu < cla hpvn;, n n y v.,.. t n. ".i . t'.- ..{.: .. ;, ; e; .-x .~ .w' y c.y - g- _. w. ' t,. y .g- !. t. t; v'y n,,15, p J g ?, g g i $.y(b i fb, - T ';yiz 4,* v's. l / ! b..f.. ^ 3 ijnt- ,3, g r 7,0-, 1 k6, -. f;i.. 5 ; 3, )Y O]/[w. g.Ig '! sCe - f U, 7 ' ]s, o i t 6 t r q . *- by <. s ,c vr. =... e' ~3 N' p > 3
~~5C ', ; v ; f' ',~~
~~) j 'pj..,.- 4f-[? mgr: f J v. ' ? . j g,, E'9 fl Fah p*co' w.p;- t ;, J i' '[ 'y u. s q. i., ,m ye t W, {'i l p:'. u 4.,~~
~~,M -[ f~~

~~nr.' f u h. ' +; ", j ], 3 e <,I. 4 S s 7 r, ,.t.3: r'.h i g%g'f., @O F.. g. y/ si a.y%.g'. r?,4 N-* r' ' sg. ' q: s xx _ p 3 5 mr. - ; J -J5 Ju p t aM.t a, s,.J * /g" e34 i f~~
~~4. t u m;..m. A.s e, s -)~~
~~g :. !, i 4 g tf e~~
~~r. yr '~~
nu i m 3 ,. c-3 v. a i .m g r y c v c-e. t -.v g 3.%. g; yu r - + i + , [y n 2h. pn:3.>s ;pp *gn ~1-3 , ; 3 g A. - ~... g 3 q y y4 et e q i j. = ' - N h, it ' Mp . n m. a., n te m 'kh iq, i 9 r g ,1 7 ,sw J M T w m, ;,. :}n a\\,;e, ,y ,... W 4>lx '..y:.,,, o, ~ , p: g< q; i g ^ s r w ,,, gp ; G ',Q.l n+w ~ q [" 'g)a
~~d t.. Q l.~~
4 s w w m'rfy s n.. p< y} '., .. p p yp un% q,n "[ [ h p f y ~1-(G. M @7 , u u -w,. vu 3 h,%MM K,w.tk M. n.. jf v NMf 4 M+dQ.-:t ^* ~ . ;: p.j m.m 4 u,U d, 4 A.,..-] ; S :y ~ w.m. %,n % =.J ~ 1 h 0 a
~~_ +.#?~~
~~m. -~~
~~= ,i ; 1 1, ~M: t Q p. ,C y,~~
k)
~~W}8 /.. tC QQ WW;, : n 9Q W b d w. g M M1 a v ,%' ' a f %f(p&Uk&p s.e&Wh W lW_k%? YrWa &~ w _cuu:k -y n L,sf ? l' 9 $.. *' l Wy#;;%_w-v a nl%,w:w%w% 0 dm u~~
~~t;wm 'm w mp.m.QqV M h ympmM :q\\,:4 W %o;mL m,.m c _vy -~~
,2 uu m,w%m#g h.gh,,#%m e&yp 9yw a, J,~ p,f ' 4. n;du r.N a #qdf g:,m.p,, e _f, m r. e w -.ds? s w%m%c, w %m, Jw n, m,,4,m.,.w., e.w e%.. < c> myiQ T.p pr s m a A. w,. m .,p% -m v g c e e mum n m m~ e.A s,. n% u
t I ~ GE Nuclear Energ;- j 1 -23A4800 . 1 '3 Reilshin I e 1 - Class I. 1 . September 1990 - i 23 A4800, Rev.1 Supplemental: Reload Licensing: Submittal-1 for. m Pilgrim Nuclear Power Station : Reload; 7-Cycle 8 = ' ~; j zi 't t I 't AL [. i g , 3 s Approved: Approved: ? Mj ,J. S. ha ey, anager . E. ansen, Manager Fuel Llc sin Reload Nuclear. Engineering _, l 3 h .l c l 1 s 2 a
Pilgrim neem Reload 7 Rev i. .q Important-Notice Regarding. Contents of This Report Please read carefully _ t 4 .4 This report was prepared by General Electric Company (GE)' solely for Boston Edison! Company (BECo) for BECo's use with the U. S. Nuclear Regulatory Commission (USNRC)' for. amending BECo's operating license of the Pilgrim Nuclear Power Station. The information contained in this report is believed by GE to be an accurate and true representation ~ of the facts' known, obtained or provided to GE at the time this report was prepared; l The only undertakings ~of GE respecting information in this document are contained ~inthe contract between BECo'and GE for fuel bundle fabrication and related services'for Pilgrlh Nuclear Power Station, and nothing contained in this document shall be construed as changing said contract.. The use of this information except as defined by said contract, or for any purpose other ~ 1 than that for which it is intended, is not authorized;" and with respect to any such un' uthorized use, a neither GE nor any of the contributors to this document makes any representation or warranty: (expressed or implied) as to the completeness, accuracy or usefulness of the informatiori contained in this document or th'at such use of such information may not infringe pr.ivately owned rights; nor' [ L do they assume any responsibility for liability or damage of any. kind which may result fr6m such use of such information. j i c
l
' {' r f' q ) i a ?y; Page2-d = 'L. ~ r
e i Pilgrim-zwsm Reload 7 Rev 1 1 1 i Acknowledgment 6 The engineering and reload licensing analyses which form the technical basis of this Supple-mental Reload Licensing Submittal, were performed by E. G. Thacker II of the Fuel Engineering, ) Section. The Supplemental Reload Licensing Submittal was prepar$d 'by P, d. Lambert 'ahd. ' .i verified by G. G. Jones of Rigulatory and Analysis Services; i _i c' r h t l i $'i r -( i -i .m Page 3 g mi m
o 'J 4 Pilgrim mam .j Reload 7 Rev 1 1._ Plant unique Items (1.0)* dl I Appendix A: Increased Core Flow Throughout the Cycle : i 2. Reload Fuel Bundles (1.0 and 2,0) Fuel Tyne. Cvele Imaded L Sumber Irradiated P8DRB282 5. 24 t P8DRB265H 6-60 i P8DRB282 6 ~ 112 BPSDRB282 7' 32' 1 P8DRB282 7 160 'New j BP8DRB300 8-192 4 Total 580-R 3. Reference Core leading Pattern (3.2.1) ' o Mwd /MT mwd /ST f Nominal previous cycle core average exposure at end of cycle:. l'8,362' 16,658 Minimum previous cycle core average exposure at end ~of cycle: 'g from cold shutdown considerations: 18,362 16,658 Assumed reload cycle core average exposure at end of cycle: .21,018, 19,067. Core loading pattern: Figure l'.
11
~~( ) refers to area of discussion in General Electric Standard Application for Reactor Fuel, 1~~
ll NEDE 240ll P A 9, September 1988; a letter'"S": preceding the number refers to the U.S. Sup-plement, NEDE 24011 P A 9-US, September 1988. 4 l; t 1 g Page 4 u
. ~ - l 4 Pilgrim; zww Reload 7 'Rev 1 4. Cajeulated Core Effective Multiplication and Control System Worth No Voldsi' 4-20 C (3.2.4.1 and 3.2.4.2) Beginning of Cycle, K, Uncontrolled ' = 1.116 1 Fully controlled 0.962 .l Strongest control rod out 0.990-1 R,. Maximum increase in cold core reactivity with exposure. .;0.000 R into cycle, AK ] 5. Standby Liquid Control System Shutdown Capability (3.2.4.3), Boron Shutdown Margin (AK)'. (pp.m) (20'C. Xenon Free) { a 675 0.040' i 6. Reload Unique GETAB AOO Analysis' initial Condition Parameters (S.2.3) t Exposure: BOC8 to BOC8+8767 mwd /MT (BOC8+7953 mwd /ST) f Fuel Peakine Factors Bundle Power ' Bundle Flow Initial 1 Dstign Legal Radial gial R Factor (MWt) , (1.000 lb/hr) MCPR BP/P8x8R 1.20 1.62 1.40 1.051-5.454 108.6: 1,44 9 '} Exposure: BOC8 + 8767 mwd /MT (BOC8 + 7953 mwd /ST) to EOC8. 1 Fuel Peakinc Factors' Bundle Power Bundle Flow Initial ~ Design Local Radial gial R. Factor (MWtt (1.000 lb/hr) MCPR f ~ BP/P8x8R 1.20 1.59 1.40' 1.051' 5.355 109.2 1.47 7. Selected Margin Improvement Options'(S.5.1) 4 Recirculation pump trip: 'No. Rod withdrawallimiter: No Thermal power mwitor: No \\ Improved scram time: No lI Exposure dependent'limi:': Yes Exposure points analyzed 2 l l i j{ Page5
u 3 Pilgrim swsm'- 1 Reload 7 Rev 1 ) 8. Operating Flexibility Options (S.S.2) j n Single loop operation: - ' Yes - load line limit: Yes : Extended load line limit: Yes + -1 Increased core flow: Yes j Flow po. int analyzed: 107.5 % j Feedwater temperature reduction: No-l ARTS Program: No Maximum extended operating domain: No i 9. Core wide AOO Analysis Results (S.2.2). a -- t Methods used: GENESIS Nuclear with GEMINI ODYN and GEXLV .. r - ACPR- , Flux. O/A ~ Event (% NBR)- (% NBR) . BP/P8x8R Eiguts q Exposure: BOC8 to BOC8+8767 mwd /MT (BOC8+7953 mwd /ST); Load rejection without 651 2125
0.32

2-bypass
-i i 1 n ~ 'f l Feedwater controller 663 132-0.37 3' 1 l failure l l- 'l 1 Exposure: BOC8+ 8767 mwd /MT (BOC8+ 7953 mwd /ST) to.EOC8-Load rejection without 730 126 0.35 4 'j bypass i iter controller .718 134 0.40 5 != .l j L l 1 I
~~i i~~
g -~ - Page 6 '
[ t. Pilgrim _ 2mam Reload 7 Rev 1;
~~10. Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary (S.2.2.1.5)~~
~~Generic Bounding Analysis Results Rod Block ACPR Reading BP/P8x8R - 104 0.13 105 0.16-2 106 0.19 107 0.22-r 108 0.28 109 0.32 110 0.36 N Setpoint selected: 110 n~~
~~11. Cycle MCPR Values (4.3.1 and S.2.2)'~~
~~Safety limit: 1.04 Non-oressurization events: H Exposure range: BOC8 _to '.EOC8~~
~~i f~~
~~BP/P8x8R Loss of feedwater heating~~
~~1.21* '~~
Fuelloading error J1.23' l Rod withdrawal error '1.40 ej s Pressurization events: j 4 Ontion A ' Ontion B': - BP/P8x8R BP/P8x8R - Exposure range: BOC8 to BOC8+8767 mwd /MT (BOC8+.7953 mwd /ST)- Load rejection without 1.42: l bypass i Feedwater controller. 1.48 ) > failure 'z
~~The minimum MCPR value required by the ECCS analysis is 1.24.~~
.l " Option B is not available for this exposure, Page 7
i Pilgrim 2msm Reload 7 Rev 1-
~~11. Cycle MCPR Values (4.3.1 and S.2.2) (continued)-~~
4 Pressurization events * .: [ Ontion A = Ootion B = q !j o BP/P8x8R BP/P8x8R i Exposure range: _ BOC8 + 8767 mwd /MT;(BOC8 + 7953 mwd /ST) to 'EOC8 L L3 Load re 1.45 1.40 ^ bypass ;jection without i) ,i Feedwater controller ~ 1.51 1,41' [ failure 'i
~~12. Overpressurization Analysis Summary (S.3)-~~
~~. Po - Pv h Event - - (pgig) (giig) Plant Resnonse MSIV closure (flux scram) 1317 1333 Figure 6 l~~
1

13. Imading Error Results (S.2.2.3.7) iq Variable water gap misoriented bundle analysis: Yes 1
~~Event ' LCP.R l Misoriented fuel bundle 0.19 k 1~~
1

14. Control Rod Drop Analysis Results (S.2.2.3.1).
l 1 Pilgrim Nuclear Power Station is a banked position withdrawal sequence plant; therefore, the control rod drop analysis is not required. NRC approval is documented in j NEDE 24011 P A US, September 1988. if .1 1 i j; Page 8 - 1
Pilgrim. zws:o Paland 7 kev 1
15. Stability Analysis Results (S.2A) g Pilgrim Nuclear Power Station is exempt from the current requirement to submit a cycle specific stability analysis as documented in the letter, C. O. Thomas (NRC) to 11. C.
~~Pieticrlen (GE), Acceptancefor Referencing of Licerning Topical Repon NEDE 24011 Rev. 6,- Amendment 8, 'Thenna! Hydraulic StabilityAmendment to GESTAR 11,~~
~~April 24,1985.~~
Pilgrim Nuclear Power Station recognizes the issuance of NRC Bulletin No. 88 07, Supplement 1, Power Oscillatioru in Boiling M'ater Reactors (BM7ts), and will comply with the recommendations contained therein.
16. loss of coolant Accident Results (S.2.5.2) f LOCA method used: SAFE /REFLOOD/ CHASTE See Loss of Coolant Accident Analysis Report for Pilgrim Nuclear Power Station, NEDO 21696, August 1977 (as amended).
i 0 0 Page 9
e in[d I, l'o BEBBBEMBEBEM ll M M B E M M E !! M M M lt MMMMMMMMMB8E
MMBEBEMBEMMMMMBEM

BEMMMMBBBEMMMBEMBE l':BEMBEBEMMMMMMMMM

BEMBEBEBBBBBEMMMMBEBE

BRMBEBEMBEMMMMMMBE
':: M M M M M M M M M M BE M M i l::MMMMMBEMMMMEEME MMBEMBEMMBEEEME M ME+sM M M M M M BEEEE+EBEEEMEE IIIIIIl-lli1III 1 3 5 7 S1113151719212325272S3133353739414345474951 4 FUEL TYPE-Bb8 SH B 82 C = P8DRB282 F = BP8DRB300 Figure 1 Reference Core Loading Pattern Page 10 .i
t Pilgrim zwuo. polnad 7 m,v i-i 'i ( sh- @!Pr" M< s % / f F -- >vi x 7 l N / \\ / I3 ...=/ 3. l 'I tow (u(M,5) gig (gggeg) iM@r"' n/ !MiMr i i J / ... N, 1" /\\/ w.- 3 s \\ g::: L .. s. sin rusam) i,ive ) i t i Figure 2 Plant Response to lead Re,lection without Bypass l. (BOC8 to BOC8v8767 mwd /MT [BOC8+7953 mwd /ST)) l-l l \\ Page 11 i I 1' l
Pilgrim mwo i Reinnd 7 kev i i i t p *Sih Mf Elr'd') lf d9k Hb'!)sI+.i rw. t,,ks E t: =. _ar _ d l C *' -m, t a i ? ra g i , 1 I
(
~~M ..a W :::: :: k a. s. tid fu;tsti) tad (Ecect) - i a I \\? 6 (f / 1 : : : a :- --s -7;=, l .g e e E F E 1 3 /O 'I ( l, t r u.. tid (ut>cl> tid (utees)~~
1.
F i l Figure 3 Plant Response to Feedwater Controller Failure I (BOC8 to BOC8+8767 mwd /MT (BOC8+7953 mwd /ST)) 1 Page 12 j 1
Pilgria mwo Reinnd 7 kev _1 o
~~4 I~~
~~'r ) m. l u,.A.,AN / 7 [ \\ / x, 3 L w .A. .r.~~

~~3.-~~

3. tid (uCMs) tid (ECDCS) r e ,1, fy[dy%lipbyhv.s.at) py'a $g**!!!!!" f V o l ^ ,\\ ,A f (' =. ..a ~g [ g p r: ~ 3 ~ l l i 1 w. i tid tutoes) tid (scoesi 't i Figure 4 Plant Response to Generator Imad Rejection without Bypass (BOC8 + 8767 mwd /MT (BOC8 + 7953 mwd /ST) to E6C8) l Page 13
8 Pilgrim mwo Reload 7 Rev 1 l tes. ) SYf$r'sY@'s1 l'fhNYI'bh"O rwa Jud: : I! i Hu !NI k s s f '" 8,, &. '= \\_ 3-b t P4 s s ...s - -- s.m. e,,, k tid (MC>cs) tid (WCMS) ag a *1,JU i $h{vfif)kau@i !,;"adaphm I i IL (m'mja r.sp. pat} 0 ( g J: - I / i v> '~ i rr
~~.. n 1~~
h n t n \\ l' ^ w. .i. l t \\ .1 i... tid tutacs) tid (aceet) 1 ) l l l4 Figure 5 Plant Response to Feedwater Controller Failure '1 l (BOC8 + 8767 mwd /MT [BOC8 +7953 mwd /ST) to EOC8) ] 1 1 Pagt 14 4 j
Pilgrim - zwn' Reinnd 7 kev 1 l t i a ) (.fI f 0 g v ~< h r * '- ~;. 8" ' a 7 b 1 \\ ~i \\ ~ j
~~3 N:~~
~~i r tid (Mttett) lid (MCDel) ~ sty em.pgr $(p.g A1) g th I hilM M ! N !!! O A l l b- - 4 g 5 s,. 6~~
4 e ae

3 % e
,e 18d (Mt>C1) tid (Etees) s I t';;ure 6 Plant Response to MSIV Closure, Flux Scram I Page 15 g t
i Pilgrim
~~me Reload 7 un. 1 l~~
Appendix A 3 Increased Core Flow Throughout Cycle 8 The analyses performed for Cycle 8 included increased core flow throughout the cycle. There are no concerns regarding reactor internals pressure drop of flow induced vibration as 1 discussed in the increased core flow analysis document for the EOC 6 (NEDO 30242). The flow blased instrumentation for the rod block monitor should be signal clipped for a setpoint of 107G, since flow rates higher than rated would otherwise result in a ACPR higher than reported for the rod withdrawal error, 4 ? ? l 9 i i I I i a Page 16 (Tenal) ' i
10CFR50,90 f ) SOSMWEW50N Ngom Nuclear F5wr staton Rock y Hil' Road l Pymouth, t/ essa:busetts 02360 ) Ralph G. Bird senet vice Pres. cent - Nsclear BECo 90-136 ~ t November 8, 1996 U.S. Nuclear Regulatory Commission Document Control Desk l' Washington, DC 20555 License DPR-35 Docket 50-293 PROPOSED TECHNICAL SPECIFICATION j CHANGE MINIMUM CRITICAL P0HER RATIO Boston Edison Company proposes the attached revisions to Appendix A of \\ Operating License DPR-35 for the Pil rim Nuclear Power Station in accordance with 10CFR50.90. The proposed revis on to the Technical Specifications upgrades the Safety Limit Minimum Critical Power Ratio and revises the Operating Limit Minimum Critical Power Ratio. l These proposed revisions will extend the use of spectral shift to maximize fuel utilization during the present fuel cycle. This change is not needed if our proposed change of August 21, 1990 is approved first. t . #A, R[. G. Bird){ (N.+~ %( RAH /njm/4820 t i Commonwealth of Massachusetts) County of Plymouth ) Then personally appeared before me, George W. Davis, who being duly sworn, did state that he is Vice President - Nuclear Administration of Boston Edison Company and that he is duly authorized to execute and file the submittal contained herein in the name and on behalf of Boston Edison Company and that the statements in said submittal are true to the best of his knowledge and ~ belief. My commission expires: VM 6" /995~ LI 14 # DATE ]~ NOTARYP0f1C Attachments: A. Description of Proposed Changes B. Replacement Technical Specification Pages C. Marked-up Technical Specification Pages D. Supplemental Reload Licensing Submitted for. Pilgrim Nuclear Power Station reload 7 Cycle 8, 23A4800 Rev. 1 I signed original and 37 copies cc: See next page l
BOSTON EDISON COMPANY U.S. Nuclear Regulatory Commission Page 2 2 cc: Mr. R. Eaton, Project Manager Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Mail Stop: 14D1 f U. S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockville Pike Rockville, MD 20852 U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Senior NRC Resident Inspector. Pilgrim Nuclear Power Station Mr. Robert M. Hallisey, Director Radiation Control Program 1 Massachusetts Department of Public Health t 150 Tremont Street, 2nd Floor l Boston, MA 02111 1 i e k 1 i i
1 i Attachment A to BECo 90-136 pescription of Proposed Chance i ProDosed ChinGes Boston Edison Company proposes to upgrade the Safety Limit Minimum Critical Power Ratio (MCPR) and revise the Operating Limit Minimum Critical Power The revision to the' Safety Limit Minimum Critical Power Ratio reflects Ratio. improved fuel designs in the core and the General Electric Boiling Water. Reactor Generic Reload fuel Application, NEDE-240ll-P. The revision to the Operating Limit Minimum Critical Power Ratio is being proposed to be consistent with the upgraded Safety Limit MCPR and to reflect spectral shift operation. Spectral shift operation-results in top peaked. flux distribution patterns toward the end of the fuel cycle. A top peaked flux distribution pattern increases the change in the Critical Power Ratio (ACPR) associated with the analyzed operational transients. The proposed revision maintains conservative operating limits. The upgraded Safety Limit MCPR and the revised Operating Limit MCPR ensure the plant will be operated safely and will not pose an undue risk to the health and safety of the public. These revisions will allow the plant to be operated more' efficiently. l Rasis for Change The NRC approved the upgraded Safety Limit MCPR in its safety evaluation for Amendment 14 to NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel", dated December 27, 1987. Incorporation of the upgraded Safety Limit MCPR is appropriate for BHR's with D-lattice fuel assemblies provided:
1) the fuel has a beginning of life R-factor greater than or equal 1 to 1.04 and consists of fuel types P8xBR 8P8x8R, GE8x8E or GE8x8E8; 2) the fuel is at least 2.80 weight percent U-<35 bundle average enrichment; and 3) the lower l
enrichment bundles in the core have operated for at least 2 cycles. l 1 The fuel assemblies presently in the core of the Pilgrim Nuclear Power Station i satisfy the conditions evaluated in NEDE-240ll-P-A.. These conditions will also be satirfied for future reloads. The Supplemental Reload Licensing Subt... ital for Pilgrim Nuclear Power Station, Reload 7 Cycle 8 submitted by BECo letter 87.081 dated May 22, 1987 has been i revised to reflect operation using a spectral shift fuel management strategy, The revised analysis also reflects an assumption that the Turbine Bypass Valves l do not open in the analysis of the feedwater Controller' Failure with maximum demand. This assumption was made because precise measurements of the. opening j time are not available. The assumption is conservative and it results.in~ a more severe transient analysis and a more conservative Operating Limit MCPR. This limit is incorporated in the proposed Technical Specification. The proposed Operating Limit MCPR will ensure the MCPR does not decrease below the proposed Safety Limit MCPR at any time during any abnormal operating transient as defined in the FSAR. i w~ -w e..e- ,e .w
Determination of No Sionificant Hazards Considerations The code of Federal Regulations (10CFR50.91) requires licensees requesting an amendment to provide an analysis, using the standards in 10CFR50.92, that determines whether a significant hazard consideration exists. The following analysis is provided in accordance with 10CiR50.91 and 10CFR50.92 for the proposed amendment to Pilgrim's Minimum Critical Power Ratio. 1. Upgraded Safety Limit MCPR A. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the proposed change to the Safety Limit MCPR does not change plant equipment, operating procedures, or governing design criteria used to protect the i plant against the initiation of any analyzed accident or used-to mitigate the consequences of any-analyzed accident, i B. The proposed change does not create the possibility of a new or dif ferent j kind of accident from any previously analyzed because the proposed change u 1 does not change plant equipment, operating procedures, or governing design criteria and the change to the Safety Limit MCPR provides the same level of j protection as the existing Safety Limit MCPR against fuel cladding failure during an abnormal operational transient.- The proposed Safety Limit MCPR therefore provides equal assurance against a release of radioactive material in excess of 10 CFR 20 limits during abnormal operational transients and a new event sequence leading to an accident is not created.. C. The following design requirement ensures an adequate safety margin is maintained: Abnormal operational transients caused by a single operator error or equipment malfunction shall be limited such that, considering uncertainties in manuficturing and monitoring the core. operating state, more than 99.97. of the fuel rods would be expected to avoid boiling transition, i The proposed change does not involve a significant reduction in the margin of safety because this design requirement, which governs fuel cladding integrity and maintains the defense-in-depth philosophy, has not changed. 2. Revised Operating Limited HCPR A. The proposed change does not involve a'significant increase in the probability or consequences of an accident previously evaluated because the proposed change to the Operating Limit MCPR does not change plant equipment, operating procedures, or. governing design criteria-used to protect the plant against the initiation of any analyzed accident or used to mitigate the consequences of any analyzed accident. 1
I. B. The proposed char.ge d. 'ot create'the possibility of a new or'different l kind of accider.t from any accident previously evaluated-because the i proposed cha.; ige does.not change plant equipment, operating = procedures, or l governing design criteria and the changes to Operating Limit MCPRs. provide. l the same level of assurance that the Safety Limit MCPR will not be exceeded during an abnormal operational transient thereby assuring a release of radioactive material in excess of 10 CFR 20 limits will not-occur during abnormal operational transients and a new event sequence leading to an i accident has not been created. C. The proposed change does not involve a significant reduction.in the margin of safety because the conservative Operating Limit MCPR ensures the most t limiting transient will not violate the Safety Limit MCPR.. Reguested schedule The requested change is not needed until'the present fuel burnup reaches 7200 MHD/T, about January 31, 1991. Additionally, approval of this-request will not be needed if our proposed change of August 21, 1990 is. approved first. i I
i Attachment B'to BECo 90-136 i List of Effective Pages Revised Pages 6 l 12 13 r 205B-2 i s i l r T ? 1 I E 2 f 1 l I t f
_1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING _ INTEGRITY-2.1 FUEL CLADDING INTEGRITY-Aeolicability: Anolicability:. Applies to the interrelated Applies to trip settings of the variables associated with fuel instruments and devices which are thermal behavior. provided to prevent the reactor system safety limits from being Obiective:
~~exceeded, i~~
To establish limits below which Obiettive: 1 the integrity of the fuel clad-l ding is preserved. To define the level of the process variables at which automatic Soecification: protective. action is initiated to prevent the fuel cladding A. Reactor Pressure >B00 osia and integrity safety limits from being Core Flow >10% of Rated exceeded. 1 The existence of a minimum Soeci ficatio.n: ] critical power ratio (MCPR) less l than 1.04 shall constitute A'. Neutron Flux Scram 1 violation of the fuel cladding l integrity safety limit. A MCPR l of 1.04 is hereinafter referred The limiting safety system trip-l to as the Safety Limit MCPR. settings shall be as specified below: ) B. Core Thermal Power Limit (Reactor Pressure 1800 psia and/or Core
~~1. Neutron Flux Trio Settinas Flow 110%)~~
a. APRM Flux' Scram Trio When the reactor pressure is Settina (Run Mode) [ 1 800 psia or core flow is lest than or equal to 10% of. rated, When the Mode Switch is in r r the steady state core thermal the RUN positioni the APRM ? power shall not exceed 25% of flux scram trip setting design thermal power. shall be: C. Power Transient S 1 58H + 62% 2 1000 The safety limit shall be assumed Where: to be exceeded when scram is known to have been accomplished S = Setting in percent of by a means other than-the rated thermal' power expected scram signal unless (1998 MHt) analyses demonstrate that the iuel cladding integrity safety H - Percent of drive flow limits defined in Specifications to produce a rated core l 1.1A and 1.1B were not exceeded flow of 69 M lb/hr. i during the actual transient. ? 6 Revision-l Amendment No. 72 t
b BASES: The statistical analysis used to determine the MCPR safety limit.is based on a model of the BHR core which simulates the process computer function. The reactor core selected for these analyses was a large 764 assembly, 251 inch reload core. Results from the large reload core analysis apply for all operating reactors for all reload cycles. including equilibrium cycles. Random Monte Carlo selections of all-i operating parameters based on the uncertainty ranget of manufacturing tolerances, uncertainties in measurement of core operating parameters, calculational uncertainties, and statistical uncertainty associated with j the critical power correlations are imposed upon the analytical i representation of the core and the resulting bundle-critical power ratios. Details of this statistical analysis are presented _in Reference-2. i B. Core Thermal Power Limit (Reactor Pressure < 800 esta or Core Flow-l < 10% of Rated) The use of the GEXL correlation is not valid for the critical. power calculations at pressures below 800 psig or core flows less than 19% of - rated. Therefore, the fuel cladding integrity safety limit is j established by other means. This is done by establishing a limiting condition of core thermal power operation with the following basis. \\ Since the pressure drop in the bypass region is essentia11y'a11 1 elevation head which is 4.56 psi the core pressure drop at low power and all flows will always be greater than 4.56 psi. Analyses show that with a flow of 28x103 lbs/hr bundle flow, bund!e pressure drop is nearly-independent of bundle power and has a valus of 3.5 psi. _Thus, the 3 bundle flow with a 4.56 psi driving head will be greater than 28x10 lbs/hr irrespective of total core flow and independent of bundle power for the range of bundle powers of concern. Full scale ATLAS test data' taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MHt. Hith the design peaking factors the 3.35 MHt bundle power corresponds to a core thermal power of more than 50%. Therefore a core thermal power 1 limit of 25% for reactor pressures below 800 psia, or core flow less than 10% is conservative, f i Amendment No. 42 12 i L!
.) i RMES: C. Power Transient j i Plant safety analyses have shown that the scrams caused by exceeding any safety setting will assure that the Safety Limit of Specification 1.lA or 1.1B will not be exceeded.. Scram times are checked periodically to assure the insertion times are adequate. The thermal j power transient resulting when a scram is accomplished other.than by the expected scram signal (e.g., scram from neutron flux following closures of the main turbine stop valves) does not-necessarily cause s fuel damage.- However, for this specification a Safety Limit violation will be assumed when a scram is only accomplished.by means of a backup feature of the plant design. The concept of not approaching a Safety Limit provided scram signals are operable is supported by the; extensive plant safety analysis. The computer provided with Pilgrim Unit I has a-sequence annunciation 4 program which will indicate the sequence in which events such as scram, r APRM trip initiation, pressure scram initiation, etc., occur. This { program also indicates when the scram setpoint is cleared. This will provide information on how long a scram condition exists and.thus provide some measure of the energy added during a transient. t D. Reactor Water Level (Shutdown Condition) During periods when.the reactor is shutdown, consideration must also be' given to water level requirements due to the effect of. decay heat. If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is' reduced. This reduction in core cooling capability could lead to elevated cladding temperatures [ and clad perforation. The core can be cooled sufficiently should the water level be reduced to two-thirds the core height. Establishment of the safety limit at 12 inches above the top of the fuel provides adequate margin. This level will be continuously monitored. References s
~~1. General Electric Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, General Electric Co. BHR Systems Department, i~~
~~November 1973 (NED0-10958).~~
~~2. General Electric Boiling Hater Reactor Generic Reload Fuel Application, NEDE-;40ll-A.~~
Revision l Amendment No. 42 13
TABLE 3'.11-1 OPERATING LIMIT HCPR VALUES l A. MCPp Operating Limit from Beginning of Cycle (BOC) to BOC + 7,953 MHD/ST. P8x8R/BP8xBR for all values of t 1.48 B. MCPR Operating Limit from BOC + 7,953 MHD/ST to End of Cycle. l P8x8R/BP8x8R t ( t10 1.41 0.0 < t 1 0.1 1.42 0.1 < t 1 0.2 1.43 O.2 < t 1 0.3 1.44 l 5 0.3 < t 1 0.4 1.45 0.4 < t 1 0.5 1.46 0.5 < t 1 0.6 1,47 0.6 < t 1 0.7 1.48 1 0.7 < t 1 0.8 1.49 0.8 < t 1 0.9 1.50 0.9 < t i 1.0 1.51 I t-6 Revision Amendment No. 108 205B-2
i'.- r / 1l _f r- .t_ -E 4_. \\% - g , i s ffgl. , l_ ff f [ Y f )ki .i f'. ff yi.k - - i ) -$ * [_M } [ _
~~' ~~~
~~I hv I ' ~ \\ ' i --{' I ? / jb'? { } f.hj ('-' 'i,,} g' z[ ' \\ \\ q.ka,. q . g'{ [-f 4 ' y-Y [ ',. }[ [- g g [ h-L _ h.. o j ( Ol S gh~~
~~t. :1.f j - ' ' - } y~~
~~'4 ,4 f { f([ ' ' m h; 4 '[: - 4N<. 3 _3 +-~ L f ,.a sJ4j jt' --Qf, P, f j -; d~~
~~4. -~~
~~3 44-~~
~~-by.!~~
~~',. Tg g i,j p,- (..;4 iy. t.1 4p'.,e 'af } b.. j~~
~~9. l.S 4~~
'? ,,a t,: ) i., 5 :. t,= ,c p3 ' V : g Y.- -- 9 V[,- tge "l .,}'... [.. i - R]*pa' )rs e 5,' r. ) t f a_. ys. g9 .th _d j '] ,4: , { w.! g g' i, , ). a ; p y g..,, u .. - z... % .fy Q '; 4+ g- ; g:rQ [ j,6',, j .z ' $ 4 j. L p f'#,- ( 7'. t e t 3 ,y+ T fx J,,,+.- 7 qg: - ~ ..,,y ,. j,,'..g '"g,. % g-d'--d - r 5 y22 4 - 5 9 ,1 = j,: I 3 . Nb 4 1-y' ) ^ h h i { }f 4 {}y 3 i INf , yh l-3I1 ,g I 4re-4,. ..4 i ) '@l '[ g s:4( ns- [ [3 ' ',-[ g ,3 1 k _a t i z y-t -j 1, ( 4 q 't t, g _ 'Q ji-l ((;ig, D!' ..[ +N ') M .y3 i fi, d 1 {- ' ] 7 g
~~[l-~~
~~[ V 4 1 {J~~
~~-,:, V ' I h-<- - : t t~~
~~.5 hi, r= [tI 'i %i F r. Y ik<8 't - 9 -1 ) y.- 4,,T I /I M j ' d_; _ 'A.; r.. = l, ' 3 4-> e... q,n]g-g.- p (= U -~~
4
~~.. s. ;~~
~~a 6~~
~~4 yi .k "4 r... 1~~
~~i n.t '~~
~~r yyyn-3 n a s.4, 3< q~~
U(
s.g;g tc 4(', t-3 i -. = _. ;. --,g, yg cj_;'t .a Q s, y 'i f 5-:p; ,pg.; 3 g { ATTACHMENT C to1MCo 90-136k f v{u sm # 1 L, t f'] > r i4W j y,9 - ! - {l k j.. t.,'f j_;'*-~ +? T.,f ',f_ -.1 b 9' k 713 ' ' ;- 1: t-- t , _ M_ jv I ( t, ,Yf, A s -_(1 ~ Narted yp Technical Specificatient :.* : 4y m n f V V,u s j '} -}",= f T. } 3, + j.' ~ _ _ 5 4,..,tc' s g e ,\\ 4l(,. _ r g * - /, 73 4.bf ' g '. f. .-l q, N4 i ,'_'i,7 g 04 .g y _s g ~ gip, s ,_c 7, d )q; ? y?.-$"i,> -I i,- - c,'- s,t i
~~v. ' IdiL~~
~~,E 9x ' ' -.,,p.. ,MY 6 i 4 t' a. u. a. '. 4.. f n [. ; ' ,'(4 g 30 i 4 a ;&k 1 if r.,,. ,y 'g , '\\g. j I ' p y' 3 3p-3 g c..- r 3 :; y, -- t,3 jyz ,.,1f-k... $ I' T-~~
~~'f' f,~~
.-f, s {p'. k m4-e . ' i[ N ~~ i .M k' 7 l'[',; y; 4 t I$,*' ':, i t J: m % 1 t ..4; s 1 i ..i ', f } 7 7.i f'. ' t -... :p ' j " *- - ( h.,,g= .t_1" 3.g-,. 7;s ye.,,Il'^.- ..,b1 _g. .. - h i. - ' [$ s ;\\ .f t 3 5s, _ ji{ il--3 Tq w n om 'mr> g., ; 4 y t
~~q 8', Y.,~~

~~4..~~
~~-,.n(~~
~~c r ;- - :, -.] } $ [ di, si }:If "$ 'a. ,..')[I g i r. w 9 _l. '= , k:* f ~, ' E.~~
~~k.c 9i,. '~~
~~4 '- I i~~
p
~~[. 4[, _f..=p.. t't.., t.Ri.y) .,. h ' j g',p! -. - Ii - ( -' i.g9ag 'j , y []$,g - P i~~
~~j ;'~~
~~+ , cf $ -.y, i E /3{%2. c5 ~ if ' c ~ 30 0 4^ 7, 2 i5 ' ( g= 'Brt t p. y~~
~~h,'i g gu g P ".~~
~~3 ,, 3 :,~~
~~p 3~~
~~-,., 4 s'._, p q t.; i , e --jg_ , a 's ,M4 w $a en .. b; 3* p. ,;'p,., f: -.f.i @;T. i 3 A, upt.. n t e g-p :; ; iJkie qi-- fl A-3 -7.g p 4 .H r1 c:%.i' P,i:. t i a b o r 5~~
~~p, e~~
~~y 4, 3i { (q 'c .t-S. . *i 4=,; 3 3. _,_ k l. p~~
.;y gg j y -_,

_y x I i C. '
~~i;j.,.,,, 1 r = Q M y * %m. L y ' q. ;;% & j,,,., %r-dqf ' 5u ,4i dt i J pi 3 f t . y ' u,, w,. i U y,w 'm'- -. .e s: a v a r ^+:? ar; + j' ' @,. m}l&, r. ep:C 1 -y a~~
h(
J ':' 4 _,'.a i.b b E e. - d2 i (, i'a. t % -. m3 2 - i V r t A .,.i _g + y!, ; ! .4 n-un ['"6... fj;f.@'.;;t [ nyj t W Fq q ~ '. q. 0 s.4 - t ( -y;T,4.-) b ' {, =}- ,y)-^ : -. ' . s W .'y' J {. g - _ %,9 ' f e,, A r i e. 'r '. j. c }J 7; fl-C h;!-..$h< y s'.1,, % i i' g ij d, e L ;i c' p$1 gg: ., '.d g,..--t.,."L ' ' M. ]e; yr n ' %:QH ,i j + j _.J ' t) }v ,,s' F w= [ r s ,sy .m@y ;. 4 u, _, -c
3 _.

am 4 g
~~+ - f q 1 ; [-} r ; 3 +v --~~
t
~~'-M Q q E t[ ~ 3V i ii! L j. t ^ I k [, y j i , y l I /-}, f,~~
(
~~[ 4 r i~~
~~t i~~
~~^: i '.- ; -:., ;?, ? I1 L' L c~~
~~4-t y~~
~~P, 4 f * -b_ ' "'h'- fj;d ;b 3 M ?-) T 7 v ' '5.~~
f
~~$1)gf',. 1. t i .6 g' $s ,/ f f f-~~
~~i. ~~~
~~jp}- a 4 71-7 y [?! b'] . 3~~
~~jirp y~~
~~,-(, fd i J d p -%, t ?' <M 19- ]e l } g y ,, ' ' w,.t... , 4~ 7 J_ c s 9x33mm q: , ti i ;_ y, < ( -_ x, a. e;s --, P, ' i ( ' 2 sp ...,s t p. - WR. 5 - of,#.1 y- ' - R @ "4 J t '~~
n.
~~st , W, r w r 4 ,.f p, 44a g 3(c s 8 ;- a 3 _u )-- A. f ' h 2-Y, Y , b.-- '. N;~~
~~K' (t '. 3 r r Iv,'~~
~~s_s -em 4~~
?
~~244.,7' Vt M, 3 r 1 i >V e ?.~~
~~A' f._ M. W. + 'Q"$.$.. ' W' H<~~
~~k' t Q .e $$ 5 1 A.. h j mo , jj f' -d Q'?A_4: i f I j- .,.j, j, t s Je 3 ',1 N i, t.,4'3',, \\$,, ; I I,, 'Y" . i;( '.,z,1 91' -.C9 ' T"9'? ? ' ( i,(.~~
~~p -,,.~~
~~-m/ fA a c 1 '..? ^ i. I ,,g', 4h' t k 4,i.~~
~~,.3.i 6 '; '~~
~~g 3 F l.= S t' r .f* g' i r ' 1 2 <;7 _=l;5 p( e i' ' " j,.;' ( e, 5 .5 T 3 )w,.- .).. ,e n, I~~
~~'f,.~~
~~s u! r . ";. $, o. ) ),.) j -, '1 I .1 ', iGi . ~ I. p: d,,4 y (; 'I 5, k L g~~
~~N. y s~~
~~s[ 1 ( <, :5 4 .l [s. ty' -,'- [ ',i l. e. ? lj, . 11, ^~~
~~A. W.g 4 i,,, p / L".~~
~~. _' l'i t c - jit ' 3 i /,,,f j,.,,. m O IVl t "Q 7 p. W~~
~~gli, Te, y'~~
~~y ,. - c > m d' / '[ 4~~
~~i t~~
t , h ;
~~t. 1t;]f,~~
'y k - 2ii + e 4, '- 1 . p-- +p--it i r1 7;. ' A l a..) 4 .f- .J. 3,i;p.. ;y+ c3 , g g r -s.. '\\ -it,: t !.4,'- '- V ,rl ., g 4 I. .ti 4 f' ,[ g A ii z g "'_ N- .i & ', '- k' 'q 'hI l-I y', .,!fI 4 ~ ; - j%;,' r R; ; f_ x 'h F y; Q 4 h.- ,, >j<
~~f f~~
~~i}7 7 1 t, ,.y. j [ f I..t. ! ;,.c a -j s,~~
\\ t t
~~. t'5 'l-y ?~~
jk
~~--- !( -~~
{o i,/. d i b,.'

'; s:
~~!I < i 1~~
~~9 g~~
~~t. -"L-g .J,. . j ?!t b, . 1 .g.g-.Jt~~
~~7 pyj e },$ g 4/,~~
~~Q ; *h h.;.';) [. s. ,, g ;3 , 'f 7. i. i -.;...-- ' r 4 . 4 t 4 t~~
~~.p,~~
~~3 4 7 ..4ig' !}< A c' t - ., t i M r.y f-.- t [. [i t, - 3 i h:' s ' f. tkD 'i 4 u. hy[ 5~~
~~I's~~
~~t s . O - j* ) 6 s- .t .' ' - i~~
~~e. y, v, i / ",i9_'y-)F 4 4~~
~~y .{ -- A '5 1 [1 6 ' } {'. .~3 j Q gi -.(,-- k"' .Ic _ 4 i r y &;f[ t,p' t _r s mf ,.;1v-t t -~~
~~a i~~
~~o i m, ). g, ' v j > f' ' [ I.A } 3 } .,6 i, 4 L. e ( f J ', j r l<, ; 5 1 s _?t 'j. '. 4 4 x%, -- t, r >+ Me a .a av ( f.' ,7 p q ~ W g~~
~~,s u yn..~~
~~4 i-- e o m g _k_ i j-b a v, s '_l4, '5',' Lb s. 4 1 J if,- g y ?.. h. i -.m m t e~~
~~^ '~~
ti, 1 - 3 g 3,A ' A .,t i g s _,. S 's n (: d y5 1. 4 a s.- _ 5 3 .t } y 5 8 + ' "h <k 4 5 bks,_ l1 I h ;' k g':m l],- i ir i s 3 i n ' n,. + ' {E 6" J.M W,,', ' 1 e x 0' { j -[.- (. I l ( n'- ? pa e g
V1 c,

a. c :t.1
~~- t_ i b u s ' G.,, ~. : b 4 f 4 g_ ' { * - [ ' ' (3 f Q-L d8 t i s j =' .h, g ~ > 4 s 3 w L l,i ,g- . W W '-Q. l k,,. j k q k .M i:y " ',,a.T L, >. s s e ,.- n >c 4~~
~~.'qg y~~
3 'g 9.h h *C .,c..t,'.* _) { 'i g k,f 4,)_ .,. jy g' '7 . e n. . p 7,$ $ m. (t' y g., # . _d g 7 n b, 1 i y_ 3.[ s' 5 a _ < i. n '.- m." 4 j f 's .g .- i, y Q t,3 __,- Q 1 + w_ i [, 2 ta1 s. - 4:.i n s
~~y..~~
~~na +.:_ ~. ^' ' l y. + ] :L... J. _ a. rr..sr g -..,';>,= -q4 : > %y . 'e'; g7 -}-[ , y q_g rn v e,~~
~~i (. a.-- y\\._~~
~~n, '^M.w,j 1, --[: s gl.,, f-._ c r n. (> r. ; p.-,'.z .!g 3 .. + + Q j3,2i# i;N, y ; o ,b h.+ M. vy y y_ . ' '-W'y s QQ@@!! $ ^ %.~~
~~V 2~~
~~i W:W' 1 r.~~
~~N, ;,~~
,9 .9x p y. a. a Tt q-7 u w f , h % @:..j, k. MQ Q, ' y fA Q %a'v}:f ( lll% a l A,Q, ':k i A e q
l i 2.1 LIMITING SATETY SYSTEM SETTING j 1.1 TUEL CLADDING INTEGRITY f2.1 TUEL CLADDING INTEGRITY f 1.1 6ATETY LIMIT j l Applicabilityt AJp_icability: Applies to trip settings of the f l Applies to the interrelated instruments and-devices which are variables associated with fuel provided to prevent the reactor s thermal behavfor. system safety limits from being exceeded.' i Objeciive: I Objective: To define the level of the pro-To establish limits below which cess variables at which automatic the integrity of the fuel clad-i protective action is initiated to j i ding is preserved, prevent the fuel cladding inte-grity safety limits from being exceeded. Speelfication: Specification: Reactor Pressure >800 psia and A. Neutron Flux Seram A. Cote Flow >101 of kated i The limiting safety system trip i The existence of a minimum settings'shall be as specified critical power ratio (MCPR) below: /,41ess thanLf8fj shall consti-l tute violation of the fuel Neutron Flux Trip Settings l. cladding integrit ty limit. A MCPR of is here-APRM Flux Scram Trip _ g inafter referred to as the' .Settina (Re.n Mode) a. Safety Limit MCPR. When the Mode Switch is B. Core Thermal Power Limit (Reatt in the RUN position, f tor Pressure 1800 psia and/or the APRM flux scram-Core Flos 1101). When the reactor pressure is S 1 58W + 62% 2 loof ( 1 800 psia or core fIow is less r than or equal to 10% of rated. l Where: the steady state core thensal i power shall not exceed 25% of S= Setting in percent of rated thermal design thermal power.- 3 power (1998 MWt) C. Power Transient { W= Percent of drive The safety limit shall be as-flow to produce a sumed to be exceeded when scram. rated core flow of is known to have been accomplished, 69 M lb/hr.. j by a means other than the expected-scram signal unless analyses _ demon- ? strate that the fuel cladding integrity safety limits defined in,! Specifications 1.1A and 1.lB were not exceeded during the actual-- 1 transient. 6 N f; t [
. _. - ~ y \\ /ADEA 7 "A " 1 l l 0; re qd..d y i i., L. s;e;i.;1cel d.1 m. iL.. e.si.11rtiss- / -lie:ed ;; *;ti; 5-1, *.;f;;r:: 3, de r P -' T-1 :: Of 9: rre j p.;r:::: listed-12 u bb b, MIe.e ;e ',,. d e.. 1.6iv. ~ / { eteenkiy F.n di n s-ib ;1e; e6. 1e T1.;ee
~~1 d " :.1 vi~~
~~-h ieis ce 3.~~
blee LM e.-d*23, hfe r

;0, r.x 2:
~~' t " -1 te c: ' rut 2 9- -tt+<~~
='
. -etseetbettene 9-: it-fc ..,_b --..a. w, 6v. . u.. w. .u y wm um. -diste tbeti;;; d =.-::: ;;hr.; :=; ;b h;h;.i;.;; Of 2: h d.;_1;. j .... w 0: S::1 f+e er rrect '--':: i: de x:: ;nx:tr: r; #;n 1: M OO 203606 :M ;M h:1; fu ;h ;ent=4=;y 1;. -. ~ i eeee*A*ttee4e-sien 1: 5109MU. "M enn dinriL:f r 1 b M ui-es-e.g it :1 4Li =:- '17 ;;;; i tit er zi 7: t z
~~;ii;;;,;ily en;; ; ged ;; e ek.ad U n din-ik.1;; L- ';;~~
~~-th p;r n: -" n f.:; - ' lie; n ;M highn har lueh. L. l wee ** di ::ibuti--- 6a Pilpi: "=1:r !=;; R:ti;; '.' ' t i dai ; ny h;l ;/:10 n 14 ;.n M.; e...ze i.Le di..16 Liv..... iu~~
~~M -eelgi;.-~~
B. Core netsal Power 11_1: (F.eacter Pressure < B00 psir er Core T1ov < 10* of F.a t ed)_ ne use of the GEIL correlation is not valid for the critical power calculations at pressures belov 800 psis or core flows less than 10% of rated. D erefort, the f uel cladding integrity g safety lir.it is established by other means. h is is done by establishing a limiti=g condition of core thermal power operation j-with the following basis. Since the pressure drop in the bypass region is essentially all elevation head which is 4.56 psi the core pressttre drop at low power and all flows v111 always be graager than 4.56 pai. Analyses show that with a flow of 2Bx10 lbs/hr bundle flow, bundle pressure drop is nearly independant of bundle povar and driving head vill be graatar than 2Bx10ge flow with a 4.56 ps1 has a value of 3.5 psi. Thus, the bund lbs/hr irrespective of total core flow and independent of bundle povar for the range of bundle powers of concern. Full scala ATLAS test data takan at pressures from 14.7 psia to 800 psia indicate that the fuel asser.bly critical power at this flow is apprortmately 3.35 MWt. With the design paaking factors,the 3.35 MWt bundle power cor. Therefore a responds to a core thermal povar of more than 50%. core tt ermal power limit of 25% for reactor pressures below 800 } psia, or core flov 12ss than 10% is conservative. s 1 l l l Anendment No. 11 I w 1 w e w , ~ - ,,.--r-s ,w e-. -en e
Insert'h' The statistical analysis used to determine the MCPR safety limit is based on a model of the BWR core which simulates the process computer function. The reactor core selected for these analyses was a large 764 assembly, 251' inch Results from the large reload core analysis apply for all reload core. operating reactors for oli reload cycles, including equilibrium cycles. Random Monte Carlo selections of all operating parameters based on the r uncertainty ranges of nanufacturing tolerances, uncertainties in measurement of core operating parameters, calculational uncertainties, and statistical uncertainty associated with the critical power correlations are imposed upon the analytical representation of the core and the resulting bundle critical power ratios. Details of this statistical analysis are presented in Reference 2. s' i
l
...= =.~ - ~.a u.:. 4 l i C. Pover Transdent Plant saf ety analyses have shown that the scrams cau sed by en-ceeding any saf ety setting vill assure that the Saf try Limit of 5:rsa times Specification 1.1A or 1.D vill not be exceeded.
~
are checked periodically to assure the insertion tLaas ara ybe thataal power transient resulting vhan a scraa adequata. is accomplished othat than by the expected scraa signal (e.g., I scram from neutron flux fonoving closures of the mata turbine -)
~~Bowever, stop valves) does not necassarily causa fuel damaga.~~
f or this specification a Saf ety Limit violation vill be assumed ' f vben a scran is only accomplished by maans of a backup feature of the plant design. The concept of not approaching a safety ti=1t provided scran signals are operable is supported by the extensive plant safety analysia. The computer provided with Pilgrim Unit 1 has a sequence annunciation prograa which will indicate the sequence in which events such as scram, ApKM trip initiation : pressure scraa i his program also indicates when the initiation, etc. occur, s eram setpoint is cle.ared. his vill provide infornation on-hov long a scram condition exists and thus provida some asasura of the anergy added during a transient. i i Raaetor Vater level (Shutdown Cendition) D. During periods when the reactor is shutdown, considaration must also be given to vatar level requirements due to the affect If reacter vatar level should drop balow the of decay hast. top of the active fuel during this tiza, the ability to cool This reduction in core cooling capability the core is reduced. could lead to elevated cladding temperaturas and clad parforation. The core can be cooled sufficiently should the vatar level be reduced to two-thirds the cara baight. Establishment of the saf ety limit at 12 inches above the top of the fuel provides adequata Thia level vill be continuously monitored. margin. L References
~~Data, General Electric Tharsal Analysis Basis (CCIA3):~~
~~1. Corralation and Design Application, General Electric Co. 3VR Systems Dapartment, November 1973 (NZ30-10958). -- W m ;;;; 0;; p t:r ?:rfc-~~
~~in1= tirn 't:2 227. Crr'nl-17 1*?d dl;;;rde C:psy 3"E-Sy+te 0:;r& t, J=0, I~~
W -000'0). General Electric Boiling Vater Reactor Generic Reload [ I E Tuel Application, KCI-24011-P. u n ) i )
~ 1 TABLE 3.11-1 ) ' OPERATING LIMIT MCPR VALUES A. MCPR Operating Limit from Beginning of Cycle (BOC) to P M + MHD/ST. P' sR/BPBxBR For all values of 1 7,ggy gj,43 ,I ] B. MCPR Operating Limit from BOC + MWD /ST to-End of Cycle. / / t PBxBR/BPBxBR t10 + / #/ 0.0 < t 1 0.1 h49- /, M2 0.1 < t 1 0.2 h/.M3 /* YY 0.2 <1 1 0.3 i h/,#8 \\ 0.3 < t1 0.4 0.4 < t 10.5 4,40- /.N (f O.5 < 11 0.6 +ree- /. 87 k / / 0.6 < 11 0.7 /.Y6 0.7 <t 1 0.8 +:43 /, M9 / s 0.8 < t 10.9 h44- / 60 f 0.9 <t1 1.0 h44- /,6~/ ,/ / I i Revisionh 205B-2. [ casd n<r.
w- ~ - _ ~, ~n..~-,- a v, ; v, m.h-- f %.& m a> m. m. m. w n eL.-.n n. mu ~ ,,1 ? g g y e, . c p, L
~~f'A n,~~
~~.., ^ ~y -. c + V ~ ? wsp t w h y,w-. g' ;,, *. %.. m p,.qs.. f,. s'sisfi' '~~ ?, l t h ? W ne s ~ .; $,i : Q :' h t sf, d t m^ " u y m. m% o m.. r _ ;Qi p: u, p$. 4 ~~~
~~m w-i n~~
o .y gy_ i @ '. y. g , ; i 4 4 <w 7 p% R. ,' y .T3pm jg;gy;. tqjy i ?f 7, l - g G, k u yg, n w egggh, .g r r N ~ >tA ',N gj 7 n- ~a m, 4 s r m c w o~, a mt. - w.m n. a 4,. ),g ii y ( o i -77 L si t, m.n '),,' az-in i %w w,a m,,1 i > .m, r 7-. u a. c- =,7 ; m n+r _ ' 9l o te l e d. o 4 *1 y. ,.~;; y c p" - 1 ' ? y. 119 $ . i F , W; r 1 i < c "e a t Ce' . i 4ty 4 3 ^ 0 7,. yp. t y=qyp.. ', ,qi. -9 y =--.of 4,.
~~s..a0..~~
~~/tu -a.. .r .ms,, c. .4wn.~~
~~M n~~
n f W.,d ATTACHMENT D to BEco 90-136 " ' ', ' + s vi . s e 7,- m .a ,va: w ..r y v, .. - a SupplementalReloadLicensing.SubmittedforjPilgriaHuclea[ Power-a i Gb JM, .t = 4.. 3, 1-Station reload;7 Cycle l8f 23A4800 Rev.L1' 7 .i ' 72 0% ;c-. E %, A* o nfNMi A, .J' Jtr 4 . +3 - - %w(.g-< i 53'.. w, w :3,.
~~mg L.~~
~~..~.1 ; ~., '.. y, ,, i -c g o. -. p . O 7 1'3 4 f ,g(. _ i j' a p,,.. F )a, s p i 6 % s y,..,., o 3 d~~
~~y. _.~~
r r -s + 4 e t. .y' h kl? 'o. ~ 1 4 m)., 5 h-a O e -, (1 r a e 'O h\\ x' s y m 4a g ;t m, i ma u 4 s.- 3 a . Q -r --QFg 'J' ,t, a y, i i W h E &;;.s J4 i i - c.,. 1 r "'w, = - g[ L 3 . f '. 8 #fD, L;' en a .- c e . I =a, s e '5rf p b L'. r, .f u s A3 U. i "m, e y,k,a ;.:F. ;t t u v. i
~~v. :, -~~
~~n r ;'.Q ' W >. aWe.- .g, =v, 'i .-,e., y(7 a \\ -). b f 's f 9^ c.. , ( f ', j". ,g~~
d
~~,E I~~
~~p';, g y, J g, 4,; ; ?~~
~~4g g _.Ae3 j, ^-( ,.a. ,,,.q r f _),7 y . f 8,' -,e-~~
~
~~t 4 ' h * *. - A ^: 'ir ypig.. i O-2 . U 4 f"-~~
~~"y A --,~~
~~4 ,S.,' ^l ~ .'t z[WI p h-l Q{ s 4 r lp ? -- 4 .3 } Q &,gy. e,d s = w ,. c. w. , ~e, [. x c - b. H, @ a m g e p,; nis'- s 1:e r i~~
3
~~,T t e o v',~~
~~-! s i.~~
e 6 n - 't r C, i g wn!( p \\.- s.:,.. t. t; f 4 r r o t ,I m,. h, t.r.a ~ ' i [Afi ,y q t g m 1 ,-y vt o ': 5f;1 i f ~ y n.s t 1 t ~-L: -+ 3. 'o. .1;' :'s, l 3,:.*4 e > ru .r s 3 s, s 1 gl s-c i _' ,4 y., + t ~ m e .9- .N. w :' ., o a.. T
~~W y~~
~~J..m <k 1 1 -m, s s 4 4' .. ww ( e. ,f 6.tl4 < 1:~~
~~)F a-,y.y~~
~~.p a 3 ., ~ w ..,ym>g-x F :,. -1 .,t. q -n i n' s I y y ->c =.n~~
~~a i,[~~
~~f,',, r_ 7.y,... _._y ? > zo( j s y' y t~~
~~s. s.,.~~
~~o. g ' ' ; th, y, 4 ,1 ,y* 3 -\\R,\\ 3 ,- m. ~., s w tk f 4 + i ' a. - f .f'y' 4 .I... i '] ? A 4 1 >(,x', , / - t {..~~
-4
~~. i e 1M L ,1 ) t i ,rq i,- e '~'}',r', { ' 7 i, +. x+ \\ (, : a 1 l h' ,p 3 ., [ .'I L ' + 1 a F .i. I 1 6 14 4 ' g. , e o. m S y1 .t .M N, s qs., o~~
i
_r ~.- -: e _ - i w, s .e ,= y., is a ,,1, -y i g --;117, p, + 'ge i .s a um p i, s vs.-v:.., p I. g'. ;,. % i, 4 3 }9 '[ g<Y. = - 4, 3 + m ,.9 -M, 1 z 'J ' us, iu .pM c " { p;f.e - v + +, ' _d- ~, 'f . i t Z v.yf f
~~a. ; ".~~
~~4-. OfC 4., F~* m, =..; ...a; ,}. e, tr , s ' my> n m;~~
~~., a 2 8, -~~
~~c. v n. s v,~~
~~n : s~~
~~- =,. r t !'^ ' - 6 't.'. 4 .P '4t'e A qi b I sg g,. p i...;,]y. 4 .;i 1 - - 3 () 's a. i.-$ ' [ 3( 3,. .-w-,~~
~~e f'~~
~~,h r /, mg4 ,8-- x c g;, n, .s >+ 4 ,v _, p.; '^~~
(.
a mi +i' g, i f, '.t,i, ',9 ',_t ig' r ,g,, 3 I h* g% '};j 9 i s. 4 y l' 'l ,e. m 4 5 IY g 3 g -'a i h, m' ,g +- g ( L_ ,I I ',~ c) y' \\' d, y 1 - A l, 1-Or g 3 4 3 1 1t' %.[, M. (M-i ' N fj s 3-II .f -* J i x -e a sw A &:l p: i M? i ,5I-L. 5 1 yk,,-_ 'd ' Y k \\ .. - _ _.('a k 1 3 4 't, . -.,.,,,,'5 f 9 N ' q' \\ g a m;' L . I T-f 2a
~~I I~~
~~I-f )'l I. a t1 - g q i r 9l ', 7 p j L [t / j, h b.' sr;; q .,4 s .f 9' s G45 ~ 15 6 g , R.; r ,J - r f ( 3* i c - u sy si s 4, v i .; t..-( k 2-f 3-T . k},( i -i.i j-7 -^ y fp [ ! --; ^ r 4 4.. D~~
)
~~.mg~~
~~i. s~~
~~. e. y _4J, 3 .s .<s A,, n; M.1 i - 4g /. s s ,, sg. .f~~
~~P a7 i~~
~~.e. _ T . 3 a D_ 1 g _d.s_, ti iI e \\ t ? s 3 g f_w ), g \\ . y h ,ej c1 'bi 3-i o f t o _ f, i -- 4~~
~~.v s~~
~~t. < j j~~
~
g,
~~g,,~~
~~4 9~~
f.
j. l :. b r g A ,,t,,'_ ,; -d "tt /.fM { 'r ' r -a .~j P'i l 4 i , j 4 Q3, ,(( 3. P g 5 I - % -[ dh e i EE f'. [j,. g E l f, t w- ~ ( ?N 6 M .i.. e ~ -} f..., c ~ + -.,., f 4a.. (.F. I-g'r 1, 4,'" g02-4p y.Q g g y 9 M t ~I.. 2V.;'+ ', = r ? nGj; . c u' f r. ,s: .y u 'y 4'. - p, s g p g' g ..f, s,, .-W-_ d7 I 5- ..g sw' ~2 s d' ---u i~J ~m' ~ r, 4 ') w;.- y s >> na ,' w ,, w c - + ..[ c 1 4 1[,,f,' _- ', h.f; -.d-.\\.
~~-d~'.f~~
~~'o_, ~ '_ h g + . 9 ' f. :' '],, ^ [+,_ a e_ w ,1 1 Md r l { o' _.._ -g ? k -.y,?- e.. mM%u _p3 'w :;1,.'v.)r./M - N-u~~
~~n m'~~
wg%g! i V. p W.?. m-.e r. i I p : ~> 3 p g%:,, +. e 6pWg v%. ; m -y - e "qf a g. w, ; g: r n s. pm +x, :w~.m.c,gw.g,L, v;A' ' -4.. Tm: Naum t,.w>- g _ +n w tv __m-.,1 p..x.n._.,n.o 3 g g p ;\\ p y .m - w;.,eQ; y, w pt , S y y.s s y
~~g M p'=" m ns g.a~~
~~, y %gg y 4,.q ,,qu_s, c.. .sq e s , * + -, q'-.'~~
~~5. $g - - -. - j_,~~
~~1' 3= -. - 3 .g a <, - e. - y.1 + 4 : -m- -pa W J 5 /. p. 3 f.. 1 r is .mmmm 3 r__ y - _. y h,,T ,"t~~
~~& L j..~~
j i 3 a
CE Nuclear Energy 23A4800 Eestslon 1 Class 1-September 1990 t 23A4800.Rev.I_ Supplemental Reload Lkonsing Submittal for f Pilgrim Nuclear Power Station Reload -7 Cycle 8 I t i l I i 1 1 r O 4 I I 47 Approved: u. %Kt Approved: .J. S. ha ey, Knager E., nsen,- Manager ~ Fuel Lie y Reload Nucleer Engineering iI l .l l-1 1 J l 'l
Pilgrim zum Reload 7 Rev i j Important Notice Regarding Contents of This Report Please read carefully-t This report was prepared by General Electric Company (GE) solely for Boston Edison. Company (BECo) for BEco's use with the U. S. Nuclear Regulatory Commission (USNRC) for - amending BECo's operating license of the Pilgrim Nuclear Power Station. The information contained in this report is believed by GE to be an accurate and true representation of the facts known, obtained or provided to OE at the time this report was prepared. The only undertakings of GE respecting information in this document are contained in the [ contract between BECo and GE for fuel bundle fabrication and related services for Pi! grim Nuclear Power Station, and nothing contained in this document shall be construed as changing said. contract. The use of this information except as defined by said contract, or for any purpose other than that for which it is intended,is not authorized; and with respect to any such unauthorized use, neither GE nor any of the contributors to this document makes any representation or warranty (expressed or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor i do they assume any responsibility for liability or damage of any kind which may result fror. such ) use of such information. i 2 r ( 1 Page _2 g
Pilgrim max Reload 7 Rev 1 i Acknowledgment The engineering and reload licensing analyses which form the technical basis of this Supple-mental Reload Licensing Submittal, were performed by E. G. Thacker II of the Fuel Engineering Section. The Supplemental Reload Licensing Submittal was prepared by P. A. Lambert and verified by G. G. Jones of Regulatory and Analysis Senices. I 4 i i hge 3 i
s# Pilgrim zws= Reload 7 Rev 1= 1. Plant unique items (1.0)* i: Appendix A: Increased Core Flow Throughout the Cycle 2 Reload Fue! Bundles (1.0 and 2.0) i iel Tyne Oveliimded Number Irradiated P8DRB282 5. 24-P8DRB265H 6-60 P8DRB282 6 112 BP8DP.B282 7 32 s l P8DRB282 7
~~160 t.~~
New BP8DRB300 8 122 1 'I l ] Total -- 580 - f l 3.- Reference Core Imading Pattern (3.2.1) 4 mwd /MT mwd /ST-i. Nominal previous cycle core average exposure at end of cycle: 18,362 16,658 - 1 Minimum pievious cycle core avera from cold shutdown considerations:ge exposure at end-of cycle - 18,362 16,658 3 Assumed reload cycle core average exposure at end of cycle: .21,018 - 19,067 Core loading pattern: . Figure 1-s t l c
~~j z~~
~~( ) refers to area of discussion in General Electric Standard App!! cation for Reactor Fuel,~~
~ NEDE-24011 P-A 9, September 1988;' a letter "S" preceding the number refers to the U.S. Sup-plement, NEDE 24011 P A 9 US, September 1988. g. Page 4 - w
ht < d'iv'.UM.dt , &g g f ;.U i.e;.g., A > i h%! ,s (t w Wh r" V Q i!c_j %.@r, y:p~@ n.3 ,., ; j . 11 -- IWM !,ilt :.j!p;M 3 x 3 jtQhgNW[m v s F 0 [2 ! ra v j flgg ii _3 ,$t3, g;. 4s l# y fg yk y m m.- pg 3 rgc rfy. y a3 W p ,e ac w
.p
~~, y 1., 4 l l $ qllu Y.,..i. mmc,- - [ ml . S .l j[h ; 4 if 'y _j s '? 8_ yy'qqow[74 , g# #(4% r yeo gamm e w m%m & g~~
~~9ngy g t 9~g;p'y qWm ;g, p.~~
~~_ yy y y;;93, m - ys - qq yjgp-pg z qq y 39; My A v,u s s m: 8 s n - Q 6.# vy o p; A w,t @ p;_ da wM2 w .j> 9_- um 1: ~ t-w - o gw w c.- i .4 @n e m~~
~~p 4~~
A, + qb ca .g pq 7 f.n V. 2 W. . w; w e g r s. Sw n ., m; ~7 gr stL o .. s e 5+.. - y +> x o m. -.Q j q% no w4 c m~, m' 4,Qt o&a Naa- - m,a s enye_.. +. g t u i1 ~ a p-. s-N i y-o t s 1 5 , &w: ~%n v ++ r, 7 . w+ my,@. v&. ;p a. y , H.. r ,.f#1 -=fh,> Q.';.r+_'-A Q' ag,q. u y; e+3-O, r ~ n g:p I - Ta } i r r.w J,v pr, '^d"-b r %, ^ ? L g)m.;; r. w-a Hn u y, mm ,9"- ,-g-m,- we i m,-. w g _ .,t t 95 3 =. 7 va p Ky y' ~'.. !+ .r-b g ' ' m'T j - 9-f,- r . r p 4 ,m. .. u L 1<
y'

4 ht my' { _-.V
~~-:w,3 e j,9 q' 7- . M-s t u d.o,. .f. sg' j 1 4 '( +. i_ .br-f, y f fi l _ f M' ?_ i -l-o g # g g' t'),,1 sd th f I I b=4 yyn zu y' wp' 'e?,[y* - hI ' p.MM%x wpl:d ;w3 @,w +y t~~
~~s. e t e ': r, p~~
~~,-Y 4 s 4 W g s ~ 7 1 J d gm.~. Whi r - Nj^3' ~ ,' 3.;m @ af; g.,.- y w m y *.' _ _n, r6-y i@) ?' we, ao + g n~~
~~2. m.~~
~~m a m' f W_ mh :1aMS 1'd D ~ ^ i'h_i a' 'I-{ ~J + .-.."/.3,. 4 - -tg ,. - E d' F i 'T g w. = 4 C 4 n'; g : s 1 4 py? h -?' [ 1 71 [- J' '7 No uu i gj.. 3 +I: 1~~
~~q h~~
~~,, g s );-jD F,h y ' @t q h'_. a#.3J,' -- i s '_.+.a. w - e v,- !' r.p {,b-I- E / 4 '.' v' _M,k > g'?w'g g~~
~~s,,~~
~~~e j t,T - E'"'^;7 'J i ' 3 ';'& } 3m; y' g;.. ; ,t_,. s1 s - u k,. o._h y q v f e y';,,;t; tm y T r.. rsi c, + - -, <- 4 dm e ~ : r L +m ,;.-A -_s-~~
~~w. r. h' s.,~~
~~d ti' T f-I _5 m.9 n z, r 3 i 4 l'r-*.. - t , H ' ' - t *,,4 n Vy. 4 4 ~ T r.Sr R 2' +fi- {< x. r - 7 CI d I:s h.!-~~
~~C v~~
~~f ~ [ I' -- d r?- + "w i ~ f, d A q '[ Gw, i $ 1 '04 r 5 4 $ * { QE ', - > p> 1 p., 4-~~
~~y5 Qaf9' h7~~
-] ~hy! y m.m.., c. 5 -< g, i c. L. p_p;, _, r, g q:m, # w, z g 4 g;; Jg $s a g ,7.39 s, 7W -.p"- 3 y 3 g(Q [%q u t q,. p"g.4 y[m[u, t s - x e u_ 4 p .n, q' 1 "y;ur ] y d, 4:c ;7 -:,-cQ >g/ ]g: _ q g s w y q ,- m 'J3) .m> %- i - i ' m, s, - 4gy u ! ny a. -. e,. k ( d.. r u wm [ g-_ -.e-m yF v.m. r.. eTu n o n-e,m u h ._c,m.'.g-.;,=.o.=i(s. 3 4 g =- = r 4 3 S y p3 7;- -- _ ) I $ k. { E g-a w w. ' -I-.yg m e
~~bN p qw i + 1 p r. L;p;3y _ <~~
~~s m 4 <ym 4, e. + t~~
~~-M; 4~~

~~6V g"h =.k- ~;.._ .W ;&, w < c/ m qx "Wt g m ;p a w = 9:Wi y m;M. w,. w-y s v. W . p: ' pg " J.,:v.y r p-r J /L 0, p < ; ;. 4~~
~~r 1 A~~
~~_Q t i 1 .--.g c ma % + :y_.~~
-m
~~,a 'g. 4 s 3 - N; h ' ' h y. ! t 1 g ,nr-un. o=m 1~~
a
~~_Qp "p :{ W g T ?, '._, ;,..., 5: p ' .y. ,. ?,. - e r Qa yr g ~, . a. 4, u_ y - )k'. q4s.. o p. '-gk 1 3, 3 Qi~~
~~wn Attg)f~~
~~,9., j...~~
~~w w!A v-n~~
~~._: c a~~
D2
~~.,_ <,; g;,__~~
~~a, 1~~
~~r zm_ .< -,i--- {". : t :: -v a .p, .7% c ..- r(p a ; t _\\a~:pn1 j Q,.f. TN [. p; $-[h:.M' fI M p- ? iu / y ?" j i j 1 i~~
4~
~~4 - E '} '., o :: ' Y ii ) g s ; h. - ' 9, 2 11g v > q,. w o e t om.e @t-n~ ', nw s n,= r = m. d.a 1 ".,~g A T~ f sg _ %. _ m-s y F--. is t .,,.s u.. + =p j $4 eg t~~
~~Q 1$~~
~~ph _ y E. 7 4 %:,m%, 3-c_; y m- -t "N ,, m Mih O - 4,[-[w -?g#* 40 s 1 t e-s a su va %~~
V A !yns',

d r
~~3_Ipip y :gf " i w[= ( m ;m%'- a31?( t 3 f .,w i w. g up t 7. &m me m,, -d. {, 3 g._ ..t .= ? N4 c. o g' y 3 my ,-.4 ,j h j { h " J ' 4 4 n3 _m, i i~~
~~3. gj~~
=-w -'4 ' j' ' j 'al A I ?$ l y-q' _/Nm; t m3g e;;.p:* cR g>%.g.. 3 4.?qM. 4, m, e - ,m-.m w E. w.Tu mM 7 y ; ' t. g/apah 4 ese. s. M,s r.- - r - i t 74- . m. ^- c m,r n r -i.- Q,,f,. j, ",, M7' p ,y j m 3 t s ,. N.o -._4, 4 q a 3 ; A gg), + x 4 y e. t =m
~~),p sq...~~
~~%);:~- c;l, e~~
~~1 4~~
~~..g j o~ 4 WV =% v % + J ' ? &w d -- d 'm. g t '. 'n t i iQ t~~
~~ilg0 gQ g m y y p,,%g g s ' q + g' W 4, a:e J n3 q,~~
~~.y, . ' 4:.y. ; .. :, - f. 4 - sy [t QVty; &.a,V +y t D, 4 'grj ? s' 4 - di M Q:, n cv 2 g p~j gg " ' p:g. n' r s,d N v m %em 4 A r= 6' ' ' f %+ wc tw 4, t i.~~
~~i,im': :, p in c~~
~~h-sp~~
~~): P' (hy, N,3, T' >th' N igc f i 5- 'A '~ fjh ' " g p-~~
~~, ?g u~~
~~.;n n~~
~~7 m & ;,~~
~~7.9 =.. 3 s a +- m b c <e x 4 s ,:n _t Q %, p.... 6 ~ s p: @m, - aL I j4 ' Np,1 ' ;-=' c<4 9Mi, y, + i e e; t T~~
~~y? =~~
~~MQ3 lAM 4 4 y >,3 : i 4 '-yg 3 r e e 94 ,; y. ,M _Mm-- t I W (kr % C: _[:r i y g s t-H u MM. 4 _., mmY - & n,?. l& ' 1 e.m$pdCQ WWpl WQ:a \\ t { #$Y % Q - W~~
~~s 3~~
~~~ w c, _,A -mu4 mw%,< - u . m m; . w r; x . 4 S>n,w,, m' _. '.u : m, t~~
~~1m i_'!( {' '~~
W ' v' i -an w ye L-1 [, 't v' J y - 2 s,. s k %. -5' s. I ' f (y 'e {y7 q g j [ .y WVQ d 1,_. i ' MM-Spy; t, d. J '.. ' " 1 ~(L aF. n & m M. %.\\ ~ ~ t i s, . n r . As l 14 a,. m- + 3 1 ~% pl ' {llj %Q y \\ ;3c. _ rj yu 2, p g;h 4' qj,! 4 - rc 'f ,'. ^ Q > +t 4 w: i
^
~~?hk> :_-< 3, s s .S s- $% f;a l _:'- {'. ,, v ; hi; ' ' N'. T',! . h- .' ~' A 'l y% T:a.-e;.,- l 1 y ! [s~\\ + 1 t I1-y> ,y + .',k i (~. ?) $..-v ,1 p, c 3. m ,96~~
i..4

m, t( h, = ( ' m su,.
~~t S@; y -9 :: 1 'N k ikNI ;~~
~~g, s l I^~~
~~Tk .m'*'r ( h, Uf ,;..g-J- k - g[4o,, % ,f p - '--_[f g:= r g ,.% g'-1 [w { y. ~,,, i., t t ,te 4 y 3.',6 j / {9'(', -i @ w q, ' l,p. P.'- - s g.q e,tt ..,x, q gyv ? j g ig: R J~~
~~gaa, j~~
~~r: s( ,..w-r 9. n f ; ,1 s 1 ,4 nm> t e. f.; b ::n,- Q y 1 m -o sU Q* > $-Q i lL ' -c n it-on 't '- f: _, 1~~
wl
~~, _ _]$ 'M s ..rsn .. m, ,y,~~
Q.
~~,v 4. Q,b, - - _ nry >, i .m p'- Ypph. 2 1 (ps, j y44 .1e~~
~~4.. -~~
~~pr,vc ^ f. :;,3 sp r .-g i 4 m,1, c, 4i .y 4_9 i e. o w . c~~
~~XfyL,~~
~~',A h L n .o e ' f-f :, -h k b Q9 g M y r.> '99a " i~~
~~h.,.~~
~~\\ 3 j WM w M^ p g'$m i .-g f' P'_ M"n ' 3 m g 1,. + 3R,' A g[w 1 + g g n 7 g pf, %..l,Q Q:~ myg ' ggF g~) g p%'. y , - 40 t~~
~~fjfm FD' y,v.M _~~
~~W N M y u ;~4 :~4 i -Qi a ;% , n =p n. v i_a ),o' a- , e N Q,p}} yw m' ~ a c e _ t f 3 +,wa x.L _1 1 :.., . :g= y p&l~~
~~m +.Lpl $. g h {,~~
~~n,. c l:sv. r w, - myg.:.,~~
~~. L~~
~~1 y % te. e 's g..Q 7,.~~
~~e, n 1~~
g u ny my m ' g.p:n n s~~ x w_ w o [ hj y ', %, W O ;[ p$)&, 4, =n -m g r Q$- % ~ ' V .Qe y_'_ %y.. s ~h r =i'L g": m at rq mpy;, 4,,49 5 & ey c inf L U ).: ,,-a-f.,.n,J W >-4;d np:rg ~ L dQ-.1 y%
~~.; c~~

~~m, et 5 ', >j ('( -M; d.t*' - h v. ; :' i q q + .m w .p_. _3 4.g j 5 s i , -) 4) ; 9. j . ~4 ;t#. b. i, .j ,.b i..., -> Ml (' " j4 - W ' ) um,, gg ( tu ' I. 1 e'm' iG .g r.w( k! y, w~~
3
~~.( 4,y + q T %[j $ t ^ e.,. e(i i~~
~~', g, "i j p3~~
~~- q,1 (&M ',, 6, s j ~i t, p,,. S k,i ,y m hh ;y ' F(n =. - t j 'f~~
~~g,-~~
~~~, -~~
g
~~< > c,; p& s ,, :gg,o/Qg 4pg gM q.ns 7, "jiNsy,h.1, , M,.p. ) i L w o" o 7 m WV t-s .mq~~
~~ax -~~
~~Q 3 m x o# _ x p (f. M, W.g, [, ym ii g N m.g$w y'v r x qd f 3 sm. ~~ L a _. w-- at s. < t . 4 m m u,.,4.y I {q v. -. s iw. n (; p.': - h , { w 14 'j,. j )3 :., q ; .t e ? d.r m - e ? ifq~~
~~n w' ',~~
{.h '; g.; -j-.-t p, .p?;QJ ;. t y-t fi - - ']q ~ P . fn,+ w' -ji t =. g.; j s .\\i g;'N[v + - n; s cy: n &g ; j,% kj.Mj .d rN'N w t4'% +3, %,(.Q.w y?< fj b f m gL31 a w + d> t '. a c.c w. w, mw dw' o + *in :m< + >v + ,t -m - '-. >= 4 -i - ', ' !n Ms .; A- - t < y-2 W'f s;h< i ....,.m. u n..'. _, tWyV) - i- ~ %- *i o..~ge,a b-4, =:1 m, ll di e i .M, - j s ti ga< 1 <p; i,, ... m[q.. cq.e e-O W. C m t. /J 4 3 ..,1.,r.3 %l:e 4 3 v s' - 3 s T y t3.p ~. t ,4 .nJ3,
~~3. i s~~
e , : gyi .g4 ,f?*- = -. ?s < t g y q : J. '7bE ,.M W;.3 q, @<Q, a Aph {, e c}q'bg n e %.f Y h m~' - t WR JG s I: - 'f j!n ? i J; . i t ,n w + . -y? - - 3 t m,. 'e n m.> A%e :,W, r* m q >+ 3 q.f s w..; 4. m~W J -c e' mi r w&r = .t._ 4 i <c4v. 94r 4 s 1 n m>nw ,<r-n,, Am q-4 m o .m , e gs 4 6 Yf d !' - i O$. -t s : p ef.? ]s.. '[, ,i EyQ r [1 af, e,.[&s'.W t.d ,. j' W u -.s
~~v. U ik.. u g. g%., n 6~~
~~,-s'gl +s , g.a < i } ' 4 1 p1 M..~.. t '.a a-m~~
e
~; ,1-m w 7 aggy s,P. WE Mg a % M'M-i p s-m '+ 4c. > ;e. . y. A,f 3 i n 4 Qf y, yn, Q,p u,. L,t :$&syem, MA G > h w m..,i,4,t gm, v.f&y, +&yy, 4 a J jh,;e $. i 9c no, + e 0 Wp M, w y .w-- g ?. pr !
~~2. Lr~w m~~
~~..&. y. v n x w n.o m,r p g . w-t y g i. ( b., 6,, 4 - go'. I b i' U %, b[.,M/ [ N M kD Si bl k sa-@ ef Oh' E, [~ a r 1~~
~~PN; N~~
+ k,'j i N.'4N.M' $#5 s M '4 q' m h ), c -- ma-a t p' m o N1W.q f, h>3 3 3 ; g 7 ,f J a- -1p # pt W., e .;,,7 s- .g f t.e,,. _, i..,y. ',' r g -T g 1'; - p gL.g,W [ ..i p MN c' ,Q ~ c n yN ;1' ' a 1,p e x}t s.} :* 'c 1 7i' M,<? +e w; .m;W : w ~.. _. .,.g g m-
~~k. _. _ f-N~~
~~.g ' f *, g 4 r i , pe. ti, m,r [ s.' 43 op t i i' ! jj .,y. 3 t, =z 3 :f i c. p ,.b' g t,-,,34' i !f, .n~,_- . #g g 3 M .l) V m, .n, @ c. e sj, i, ~ y,, m. f 'O x ~ r n. r r . ~V f, = : g' =" -,c~~
~~, m n,~~
i u'] 6 ?j,9f 4 Nt' t I.W. + 1g 51, L. ?a - e j, e j ?' S x 44 4 y: n.y b 3 s .r m 9-e-U w, 4 a,, g3. s ...f '. m :'U;N. %' I _ tb'. j f. - 6,I J p .a -.,i s.. I e -4 b f_. sPe -a mu 3 .p I .Ow ~.VO.y :q w i-u i %y s m. n, A.. i n,-
~~r - a gw,~~
~~+hbhM 'Y ., j.., i [._ .7t > J-...3 . m r o' i. a.,. y ,.,. ' :. p w% + J( . -q' I. .t [.. . [ k 1,4 f 1 1..* -, =. y .' c. J v,+~~
~~d.,. h-s.. 'n.f. ;t @w. (W.~~
~~6 4 @f.%m,a. W hw' n : '. r i.p;d "ly s, 2 tQs Q; D ? T;; A n~~
~~n. g 4_v~~
~~c-g%fS F~~
~~E u^a 1, 1 gg g .,,7 p eis~, yit n,o e 16h -fm,.y rm .e u.m, mJ o,m u%. ,n, y o g W. i. ,4 o s e% e _Op% D w,.< w +w wL#v W Qbw. -ga%m-# #s w > a w-w w., m +s = a 9_4 s juPd. t rts~~
~~> * $n. m~~
~~,;CH:cN n a 4 y%_ m f>:94f;5i. au;Q Q MP , - 1; E, l psy6 @3 %,Wy- @) e>v .o 4 P: .k 5h lM e~~
~~m{,[,wQ Qwl,6 >,n N;r$yyts# ' f jQQ:;..'= Q' f:thz a~~
n ^m.. w;l#y%,'L g%, % ~ a&mrm Q?i@q p L $ _ QW:em bQ i o ., y-h i p .j qO 4., Nhh b. [h h$. h [,. ~v %s w;gm4 i o m q o # [N h hb h h N. I *, ifhNh
I ' Piigdm 2wsm Reload 7 Reil 4. CaQculated Core Etrective Multiplication and Control System Worth'. No ' olds, t 20 C (3.2.4.1 and 3.2.4.2) Beginning of Cycle, K,y Uncontrolled - 1.116 1-Fully controlled ' O.9621 Strongest control rod out 0.990 R, Maximum increase in cold core reactivity with exposure p into cycle, AK O.000 5. Standby Liquid Control System Shutdown Capability.(3.2.4.3) ~1 Boron Shutdown Margin (AK) s (pgm) - (20'C. Xenon Free) -4 675 0.040 - 6. Reload Unique GETAB 00 Analysis Inlllal Condition Parameters (S.2.3) Exposure: BOC8 to BOC8+ 8767 mwd /MT (BOC8+ 7953 mwd /ST). Fuel Peakine Factors Bundli Power ' Bundle Flow Initial: Design Legal Radial Mial
~~R. Factor~~
- (MWtr 1(1.000 lb/hr) MCPR BP/P8x8R 1.20-1.62 1.40 1.051 5.454 '108.6 -1,44 ci Exposure: BOC8 + 8767 mwd /MT (BOC8 + 7953 mwd /ST) to EOC8. -{ Fuel Peakinc Factors Bundle Power. Bundle Flow Initial. 1 Design Losal Radial Mial R Factor. -(MWt) (1.000 lb/hr) ~ MCPR l i BP/P8x8R 1.20 1.59 1.40 1.051 5.355 109.2 1,47 7. Selected Margin Improvement Options (S.S.1) 1 . Recirculation pump trip: No' Rod withdrawallimiter: No Thermal power monitor: No Improved scram time: No I Exposure dependent limits: Yes 1 Exposure points analyzed: 2 1 I-Page5-' .] q l
4 i Pilgrim mem Reload 7 Rev 1 i 8. Operating Flexibility Options (S.5.2)- -{ Single loop operation: Yes. Load line limit: Yes Extended load line limit:' Yes s Increased core Dow: Yes l Flow point analyzed: 107.57e ' Feedwater temperature reduction: No ARTS Program: No l Maximum extended operating domain: No 9. Core wide AOO Analysis Results (S.2.2) 4 j Methods used: GENESIS Nuclear with GEMINI ODYN and GEXL ,_ACPR-Flux O/A Event (% NBR)- (% NBR) ' BP/P8x8R. Eigur.c } Exposure: BOC8 to BOC8+8767 mwd /MT (BOC8+7953 mwd /ST) ~ Load rejection without 651 -125 0.32. 2 bypass l. Feedwater controller 663 132
0.37

3 l
failure 1-l j Exposure: BOC8+8767 mwd /MT (BOC8+7953 mwd /ST).to EOC81 Load rejection without 730 126;' .0.35 4 bypass 1 Feedwater controller '718 134: 0.40 5 ( failure l lj i .i 4 Page 6 y; n r
4 d Pilgrim; 2wsm Reload 7 Rev 1 1 10.- Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary -l (S.2.2.1.5) Generic Bounding Analysis Results i Rod Block ACPR ~ Readine BP/P8x8R-- 104 0.13 105 0.16 106 0.19 107 0.22. 108 0.28 109 O.32 > .i 110 0.36 I Setpoint selected: 110 m
~~11. Cycle MCPR Values (4.3.1 and S.2.2) -~~
~~y i Safety limit: 1.04 Non-nressurization events: Exposure range: BOC8 to EOC8 .i BP/P8x8R I Loss of feedwater heating~~
1.21*
Fuelloading error ' 1.23 * ' Rod withdrawal error .1.40 - 1 Pressurization events: Ontion A - Ontion B- 'BP/P8x8R BP/P8x8R .I Exposure range: BOC8 to BOC8+8767 mwd /MT (BOC8+7953 mwd /ST)- Load rejection without 1,42 1 bypass - Feedwater controller '1.48 failure
~~The minimum MCPR value required by the ECCS analysis is 1.h4..~~
" Option B is not available for this exposure. Page 7. g s , - - ~. ..L.-. ~ - -
I 1 '[ c Pilgrim 2wsm - ' Reload 7 Rev. 1 11..- Cycle MCPR Values (4.3.1 and S.2.2) (continued) 4-Pressurization events: .] Ontion A Ontion B BP/P8x8R : ' BP/P8x8R v Exposure range: - BOC8 + 8767 mwd /MT (BOC8+ 7953 mwd /ST) to-EOC8 j 1xad rejection without 1.45-1.40 bypass Feedwater controller 1.51 1.41 failure
~~12. Overpressurization Analysis Summary (S.3)~~
~~Psi Pv Event (psig) (gsig) - Elant Resnonse L 131h 1333 Figure 6 MSIV closure (Dux scram)- t~~
~~13. Imading Error Results (S.2.2.3.7)~~
~~Variable water gap misoriented bundle analysis: Yes Event ACPR Misoriented fuel bundle - 0.19~~
~~14. Control Rod Drop Analysis Results (S.2.2.3.1);~~
Pilgrim Nuclear Power Station is a banked position withdrawal sequence plant; therefore, the _ control ro'd drop analysis is not required. NRC approval.is documented'in NEDE 24011 P A US, September 1988. l 1 t l 1 3 1- ' Page 8 .i
~~3-et 4~~
i r
s
~~Pilgrim -~~
mam Reload 7 ~ ~ Rev 1-c 15. Stability Analysis Results (S.2A) - l g ) t Pilgrim Nuclear Power Station-l's exempt from the current requirement to submit a cycle specific stabuity analysis as documented in the letter, C. O. Thomas (NRC) to H,- C, q Pfefferlen (GE), Acceptance for Referencing of Licerning Topical Report NEDE 24011 Rev, 6, Amendment 8, ThermalHyriraulic StabilityAmendment to GESTAR H,' April 24,1985. Pilgrim Nuclear Power Station recogrkizes the' issuance of NRC Bulletin No. 88 07, Supplement 1 Power Oscillatioru in Boiling Water Reactors (BWRs), and will comply with the i recommendations contained therein.
~~16. lass-of coolant Accident Results (S.2.5.2);~~
i LOCA method used: ' SAFE /REFLOOD/ CHASTE See Loss of Coolant Accident Analysis Report for Pilgrim Nuclear Power Stahlon,' NEDO 21696. August 1977 (as amended), m g. - [ -[ - i e t-Page 9 e
4 Pilgrim uusm- ' Reinad 7 ked. .I Q f @+@d@@se_@o@+@@te+@o@_m 4 52 00 s oe s o So YYY d [N l 4
JEMMMMMMMMMM
MMMMMMME+EMMMMM
~
~~M M M M M M M M M M M M M~~
~~':MMMMMMMMMMMMM ':: M M M M M M M B E M M M M E ';: M M M M M M M E E M M M M M '::MMMMMMMBEE+EEEEEME 1~~
~~E+EMMMMMMMMBEEEME i~~
BiM M M M M M MEEM E -@+@ 0+$ $ _8 0+0 Q_E Q,0 b-0 0iE Q,@: 8 @ m E s e e n n e s s e o emTe o @ s 00 E E @Y@ 00E 00 ciE @; 00@@@@@@ 4 @Y@[IliccE Y 2 l' l l :ll.l-l ll.i.l'l'l.l ll 1 3 5 7'9111315171921232527293133353739414345474951' Ji FUEL. TYPE. A = P8DRB282 0 - P8DRB282 -1 B = P80RB265H -E BP8DRB282' a .C.- P80RB282 F = BP80RB300 Figure 1 ~ Reference Core Imading Pattern l I l Page 10 (9 i L L =
1-i r, ~ Pilgrim
~~naam~~
~~. Reload */ Rev 1= M 5 + i 1 !IdMdMM'"" !M'$!ll['" r f* 5,..k.,y% / i I \\ m,'w /, N l 2 0 20 ~ i 4.- ' l i. 2 .i - 21. . s.. tid (M:305) . tad (scaes) -l W >v w .. l-A~~
~~s. af~~
~~&=:- ii E e-w,,_, 3 i ) .i~~
~

~~) =*. >s.. IlW (M:Oct) tid (ucoes) - j ) ll s i~~
~~1 4-Figure 2 Plant Response to Imad Rejection without Bypass (BOC8 to BOC8+8767 mwd /MT [BOC8+7953 mwd /ST)).~~
,l 'd Page 11 lt'. + w r-e e
l V l 8 Pilgrim mom Reload 7 Rev 1 1 t %..M!! ?ti@YSP'I Iffis/Mksi rivi He isl ;P ltt-I,\\ l, ac'i a 1 1 m ra \\ l s 4 ..,fl ( g tid (MCMS) i tid (ECMS) L VQ'ji' j' h@fff 8f"-#sI%h[f -[ lQ f S kRI) Y deudp n g y m i if / =- ., l-1 i e 4 n 3 .l.. if3 /\\'Y t l .7 i... l
~~d tucoch tid (acces) l l~~
~~l-, l t Figure 3 Plant Response to Feedwater Controller Failure - (BOC8 to BOC8+8767 mwd /MT [BOC8+7953 mwd /ST))' I; 1 ' Page 12 t 9 i e r + -- +~~
~~il Pilgrim -~~
i 2msm - Reload 7 nev 17 s _e ,.i 4. j h' >lYs$r'aNAt arf rtes . h TIN EM It' I I Yetat roui - ) < hug v a w p w-3 [ i e,...k, W- / I N ~.. N, =. 4 ./L er. .x. - - -,. - a. tid (stCCs01) tid (ECMS) -l l{ -phrSEF.SiRT)
~~1. M... e@a:i!!,I,?"~~
~~11 v41v }f)v{'(so 1 g,f 5~~
k.
~ m ' Y: i .R / L m.~ l..: y y 2\\ ( y g: .r 7 W /- 5 W::'i :* : d l l 4 \\ ~s. t. s%. s. t#w (secus) tid (ucxs> - c P 3 1 8 '- Figure 4 Plant Response to Generator Lead Rejection without Bypass (BOC8 +8767 mwd /MT [BOC8+7953 mwd /ST) to EOC8). I i l . Page 131 >f l0-
Pilgrim - i m4sm Reload 7 Rev 1-i A i 4 a AfyJ15f),'EW4 Yht rt u. Pfdd)"df fd!'"i F
P

s Mis ;g itt; li
~~-t .../\\ sr - C w s,,- t E ' r-a e \\. 4 i = ..i,,.,_ .s s --.s L i tid (Etecs) Ind (Etec5) P igqppgr siwat) i 4v g agn a 1 !hfi'M:b.^d';E' !RP!fdh II i 3 1- / -= E1~~
4 1

- :
g,... y. a ]l' e ly h... }?b a i 1 i l-or j - tid (ucacs) tid.(a coes)- 1 ll.. 1 i 1. Figure 5. Plant Response to Feedwater Contmller Failure - (BOC8 +8767 mwd /MT [BOC8 +7953 mwd /ST) to EOC8) i t Page 14 1 . 1-t j)-
O Pilgrim: j u x4soce Reload 7 Rev 1-c; ,i-j 1 A> n t LPPI! ( (P14 ). .t L.R
~~(At Flux--~~
~~(( L4v '{ 1%4v v.= l. '3C C IAE FLOW =' ,m m. 4 h h 1 E, 3f% i 6 .N =. y l - -. l;. r. er.~~
~~s w..~~
~~s. Ild (MC&CS) Ild (MCDC$) - i !l t iLgvj~~
~~w.ejsgsto.yn1)~~
'gvigdtivity Ndf[gd(so Edd* P9T!!P !!N!!U j -i jl _A p.... ...~ u r 8 3 p k,, I 1 .. r. s%. tid cucacs) tid (ucoes) 'l 8 .i l - Figure 6 Plant Response to MSIV Closure, Flux Scram - f l lt Page 15 - 2 a l +
~~v...~~
o Pilgrim . mum Reload 7 Rev'i i ~ Appendix A. Increased Core Flow Throughout Cycle 8 i i The analyses performed for Cycle 8_ included increased core flow throughout the cycle. l There are no concerns regarding reactor internals pressure drop ~of flow induced vibration as - discussed in the increased core flow analysis document for the EOC 6 (NEDO 30242); l The flow biased instrumentation for the rod block monitor should be signal' clipped for a-setpoint of 107%, since flow rates higher than rated would otherwise result in a ACPR higher than : reported for the rod withdrawal error. .I q ~b ~! i { H i 1' 1-A ' 1 :- /, . Page 16' 'I 1 ' (Final)L i 1 .i -.}}

ML20062E365

Text

Navigation menu

Search