ML20127M857

Safety Evaluation Accepting Use of 8x8 Fuel

ML20127M857
Person / Time
Site:	Monticello
Issue date:	05/14/1974
From:	James Shea, Ziemann D US ATOMIC ENERGY COMMISSION (AEC)
To:
Shared Package
ML20127M816	List: ML20127M813 ML20127M830 ML20127M841 ML20127M857 ML20127M861
References
NUDOCS 9211300492
Download: ML20127M857 (10)

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SAFETY EVALUATION BY THE DIRECTORATE OF LICENSII;G SUPPORTING AMEND:E::T NO. 3 TO LICE::SE NO. DPR-22 (Cit't'!GE NO.14 TO APPEliDIX A TECHNICAL SPECIFICATIONS)

NO2THER'i STATES PO*?ER C0"?AW MONTICELLO MUCLEAR GENERATING PLANT DOCKET No. 50-263 INTRODUCTION In letters dated November 19, 1973, as supplemented by filings dated Decemb er 14, 1973, January 15, 1974, February 8, 27, and 28,1974, and April 1,1974, Northern States Power (NSP) reques ted authorization to operate the Monticello Nuclear Generating Plant using a partial loading of 8x8 fuel, including a fuel assembly containing segmented test rods, and also proposed changes to the Technical Specifications related to limiting conditions for operatien associated with fuel densification for the Sx8 and 7x7 fuels.

Tne use of 3x8 fuel in reloads has been reviewed on a generic basis by the Licensing s taff and the Advisory Com=lttee on nonctor Safeguards (ACRS). The reports based on these reviews uere transmitted to MSP by letters dated February 11 and 20,1974. 'Ihe staff Safety Evaluation for the use of 8x8 fuel assemblics in the Monticello facility was transmitted to NSP by letter dated April 8,1974.

NSP reques ted review of changes in lettera dated January 23, 1974 and March 1,1974, as supnlemented by filings dated March 8 and 19 and April 10 and 26, 1974, relating to pressure relief, control rod scrau times , s tandby gas treatrent sys tem and reactor vess el temperature r.? as urece n ts . By letter dated April 10, 1974, NSP requested approval o f propos ed changes (1, '4) that had been previously described (11,12.13) la ;ime to raturn to ;muer cp?tatica on May 5,1974. A p reliminary response to our April 4,1974 letter on the Prompt Relief Trip System (8) was submitted by NSP letter dated April 26, 1974.

Uc have previously reviewed (7) and approved (10) the 8x8 fuel rod array l

including a fuel assembly containina segmented test rods as " reload 2" fuel for the Monticello core and authorized (9) insertion of the Sx8 i

reload-2 fuel assemblics into the Monticello core for reactor tes ts .

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.t up to 1% power level (5) . In addition, we have met with NSP representatives j on two separate occasions (2,6) to revieu reactor operation during 3 Monticello fuel cycle 3 with nnd without reliance on thu proposed prompt relief trip (1,4) sys tem (PRT) to. preserve acceptable design margins to a fuel damage threshold (i.e., MCHFR > 1).

i In response to the MSP proposal (1.4) to install a prompt relief sys tem j (PRT) "to compensate for equilibrium core scram reactivity insertion-

] functions by minimizing the peak pressure and fuel thermal effects resulting f rom pressurization type abnormal transients", we stated (2) that the Directorate of Licensing safety review of the proposed PRT

! system will be completed by September 1974. W requested (8) additional 4

information related to design and installation of the PRT system to assure that existing reactor protection sys tems t.nd engineered safety-i features (auto-depressurization) were not affected by the new PRT g system connections or by activation of the PRT system af ter completion of the modification. Because we have been unable to resolve our expressed

concerns and since NSP has been unable to complete, at this time, a response to our reques t(8) for information related to the' potential for

, chort circuits that could provide power to the safety / relief valve l solenoids, thus opening all six of the valves resulting in unintentional i

primary system blowdown-depressurization, the PRT panel will not be

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connected (14). The PRT system will, therefore, remain inactive (14)

! when the plant returns to power production after refueling and completion

! of the plant modifications to connect the new safety / relief valves ,

(replacing the four safety valves) to the torus suppression pool.

The status cf the PRT " hook-up" has been reported (14) by NSP, and we

concur that plant safety margins have not been reduced by the wiring i between relief valve solenoids and drywell penetration. Accordingly,

, the PRT system is eliminated from further evaluation at this time

! pending resolution of the potential for power shorts that could open

( all six safety / relief valves.

j The considerations involving . changes to the Technical Specifications

' before normal resctor operation can be resumed (with the proposed PRT system (l) inactive) are lis ted below in the order of decreasing importanca.

l. Also included are tuo items (analysis of abnormal core transients and i ini densifica:lon consicaracions) whirh wara not addresced in our Safety Evaluation dated April 8,1974, on the subject of operation of

the Monticello facility with 8x8 fuel.

1. transient analysis, safety / relief set points, steam relief '

flow - capacity, and primary coolant system boundary stresses (proposed (13) changes 4, 5, 18, 19b, 20a, 20b).

I 2. control rod scram times (proposed (13) change -16) .

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3. fuel densification (proposed (12) changes 1, 2, 3, and 4),

4 l 4. Ex8 fuel rod array, reload 2 fuel assemblies (proposed (ll)

, changes 1, 2, 3, 4, 5, 6, 7, and 8) ,

i j 5. s tandby gas treatment sys tem (proposed (13) change 24) .

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! 6. reactor vessel temperature measurements (proposed (13) change j 19a).

i l Each of the above iteta is considered in the following evaluation.

j EVALUATION i

j Safety / Relief Valves i

j The reactor pressure relief sys tem limits overpressurization of the-j reactor coolant sys tem to prevent failure _of the reactor vessel, piping, j or components due to excessive coolant pressure.

l Prior to the April 1974 plant outage for core refueling, excessive

stress in die reactor vessel and core coolant recirculation system 4

due to transient pressure increases was prevented by automatic opening I

of as many of the four safety / relief valves and four spring-loaded safety valves as necessary whenever coolant pressure exceeded safety / relief

set points (801080 psig and 1240 psig) . High pressure reactor scram

! uas set at 1075 psig and remaina unchanged for fuel cycle 3. Each .

l of the safety / relief valves was set to relieve steam pressure when i pressure increased to a nominal value of 1080 psig. Each safety / relief L valve when fully. opened relieves M 800,000- pounds of steam per hour. ,

! Prior to April 1974, if the pressure transient exceeded the relief

, flow capacity of these valves and reached 1240_ psig, as many of the four saf ety valves opened as necessary to limit the 'pcak transient l pressure to acceptable-levels. Each of these valves , when completely j opened, could relieve 642,000 pounds of s team.

Closure of all main steam isolation valves (MSlVs) uith dalayed reactor

- seren f rom the high neutron flux nignal and end of cycle (EOC) 2 s' cram i

reactivity assuaptions was the ' oasis for determining the pressure relief flow capacity requirements. The peak calculated. pressure (at the-i bottum of the reactor vessel) for this ' assumed condition was 1308 psig or 67 psi below the maximum overpressure design limit of 1375 psig.

, Whenever the safety valves opened, s team was relieved into the contain-ment drywell causing pressurization. Since the period of core coolant

! _ overpressure is normally of short duration, the safety valves reclosed within seconds and the resultant containment pressure increases were l

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very small and well within the containment design capability. _ The unanticipated release of steam into the containment drywell through

premature opening of safety valves has occurred several times at Monticello and procpted technical specification changes to increase safety valve set points (15) to reduce the probability of unnecessary valve opening.

3 To eliminate this safety valve potential for unnecessary containment '

! pressurization, the four spring-loaded saf ety valves were replaced

! during the April 1974 outage with four target rock safety / relief valves i equivalent to the four that have been installed at Monticello since j the plant startup. The safety / relief valves- have higher relieving j capacity than, the safety valves. Also, new piping, which does not 1 adversely af fect the safety function of the safety / relief valve sys tem, was installed to deliver steam relieved through the valves to the torus suppression pool (four new 10" lines similar to those provided for

. the originally installed safety / relief valves) . The modifications were designed in accordance uith the same code requirements to which i

the original plant was designed. The modifications do not involve a j change in the technical specifications.

! General Electric has compared the analytical results(1) for the worst-i abnormal transient, turbine trip without turbine bypass, and delayed scram due to high neutroa flux with the calculated results from the

, assumed closure of all four M31Vs and delayed scram due to high neutron

flun to determine the total required safety / relief valve capacity.

i Based on these calculations, the assumed isolation of MSIVs resulta

in the greatest demand for steam relief capacity and continues, as

before, as the basis - for ' determining ~ steam relief capacity requirements.

I l For the most conservatively limiting conditions at the EOC 3. CE has

! calculated (l) that the peak transient pressure at the bottom of the '.

l reactor vessel with only six safety / relief valves (set 1 points 1080 psig) l 13 1285 psig. _ This is noticeably lower than the 1308 psig calculated value for the limiting period at the EOC 2 operation jus t prior to the April 1974 modifications when four safety / relief (set point 1080 psig)

- and four safety valves ~ (set point 1240 psig) were required for over-l pressure protection. Since the calculational methods are unchanged (l),

i S5? has proposed (13) that a minimum of six saf ety/ relief . valves be

! operative during normal reactor operation. We have discussed (2) this ,

l proposed reduction in pressure relief capacity with NSP _ representatives.

Six safety / relief valves can relieve 71% of nuclear boiler rated steam

[ flow (1) compared with pre-modification capability of about 54% flow through four safety / relief valves and about 38% through .four safety m v . -- ., e or=s.--,. #.wwE- % m ,i.e m . ,.rr-.ye----.,

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l valves for a total of about 92% and have concluded (3) that such a reduction cannot be approved before we have completed our evaluation 3

of the calculational cethods(17,2,3) used by GC. Additional infor-i mation provided in ESP letter dated April 10,1974, shown that therc

{ is sufficient margin to core coolant boundary and nuclear fuel f ailure thresholds for the abnormal turbine trip without bypass transient with

! si:, safety / relief valvea since the maximum transient presaurc during I the first 4200 mwd /T of cycle 3 fuel depletion is 1192 psig (compared with 1375 psig limit) and the lowes t critical heat flun ratio in the -

uame period is 1.68 (compared with a MCllFR limit of 1.0) . For the 1 period beyond 4200 mwd /T during cycle 3 f uel depletion, the calculated j margins to f ailure thresholds are acceptable uith steady state power level restricted below 95%. We have changed the proposed technical

, specification with NSP telephone concurrence (April 24,1974), to j require a minimum of seven operative safety /rclief valves instead i of the six proposed by NSP because at the same 1240 psig pressure where steam flou capacity with four safety and four safety / relief valves was calculated to be about 92% of the 100% rated steam flow capacity, seven safety relief valves provide about 95% a slight improvement over the core coolant pressure relief capability prior to April 1974. This

{ change is considered to be an interim change until we have completed j our review of the GE calculational cethods. We have concluded, therefore, that coolant pressure relief capability is unchanged by the conversion l f rom safety to safety / relief valves and that abnormal transient core behavior with Sr3 neload 2 fuel assemblics in the core as well as 7n7 f uel assemblics is acceptable and consis tent uith earlier core per-f ormance characteris tics . We also concur that the PRT system connections that have been made so f ar do not af fect existing engineered safety i features or reduce reactor safety because they have not been energized and thus serve no f unction.

3 Gontrol Rod Scram Time

] Ue have reviewed and previously approved (16) an increase .in the required l rate of insertion for the first half of control rod insertion travel following a reactor scram signal. Reactor operating experience has shown that actual nessured control rod scram times are significantly 1:sa than the performance limits that had been previots1v es tablished I

cnc that the avecade rod ceram Limo for 90% insertien can oe reduced from 5.00 to_3.50 seconda. Also, _ the time for the three fastest control j rods of all _ groups _ of _ four control _in two by two arrays can be reduced to "no greater than 3.80 seconds at 90% of the rod length inserted i

instead of 5.33 seconda". The reduced control rod scram times, based on reactor operating experience to date, are sufficiently longer than measured control rod scram times to allos for normal changes in control rod performance without reaching the technical specification limits.

-The original scram time limits were conservatively specified to allow for- uncertainties related - to control rod scram time deterioration.

i There are now sufficient control rod scram time measurements from operating reactors to reduce the-allowance for uncertainties.

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Calculated scram reactivity is based in part on the technical specifi-cation scraa time limits and using the new more stringent technical specification scram times specified, the calculated scram reactivity for the last half of rod insertion is somewhat faster and the uagnitude of the power mismatch and pressure increase when the normal heat aink is unavailabic is reduced; i.e. , the calculated consequences of abnormal transients are less severe because control rods are assumed to scram at technical specification values.

Fuel Densification - LHCR Changes to the General Electric fuel densification model have been reviewed (19) by the Directorate of Licensing on a generic basis with the conclusica that the Get.eral Electric calculational model,as modified by the Regulatory staf f, is suitably conservative for the evaluation of densification ef fects in BWR fuel, where possible densifi-cation ef fects are listed as:

1. power spikes due to axial gap conductance,

2. increase in linear heat generation rate (LUCR) due to pellet length shortening,

3. creep collapse of the. cladding due to axial gap formation, and

4. changes in stored energy due to increased radial gap size.

Since the assembly average -s tored energy is one of the important inputs to EUR LOCA evaluations, a technical specification limit on maximum permitted assembly power is imposed. This requirement is satisfied by maximum average planar LHGR restrictions for initial, reload 1, and reload 2 type fuel assemblies which constitute the Monticello core during fuel cycle 3 operation. The calculated peak clad _ temperatures do not exceed 2300*F with the proposed MAPLHCR technical specifications which we have concluded are acceptable. Although the power produced i by the 8xS reload 2 fuel rods is reduced (assuming fixed fuel bundic power), on the average, to 49/63 of 78% of the f uel rod poser in the 7x7 fuel assechlies, the average heat transfer surface af the 833_ f uel assemblies is only ^ increased by 13%. According to CE thermal-hydraulic calculations, the modest improvement in heat transfer capability is nearly offset by the reduced flow due to increased flow resistance.

To maintain minimum critical heat flux ratios above 1.9, as in the pas t , the technical specifications mustlbe changed to limit 8x8 f uel rod LEGR to 13.4 kW/f t where the equivalent 7x7 fuel rod LHCR is 17.5 kW/ft. We concur that these changes in technical specifications v411 maintain equivalent margins to the fuel design limit (MCHFR > 1) and fuel damage thresholds following design basis loss-of-coolant accidents.

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i 8x8 Reload Fuel i

Khere only 7x7 fuel assembly characteristics were identified, it is

necessary to change the Technical Speelfications to include the 8x8 +

i fuel asse=bly characteris tics. The changes include calculation of j- reactor scram and rod block set points based on peak heat flux in the j 838 fuel assembly rods. The 7x7 equations remain unchanged.

A new relationship of peak fuel rod heat flux versus reactor power

] level is included in the Technical Specifications to show the

, relationship of the Sx8 fuel rods as well as the 7x7 fuel rod that l had been included previously. The ax8 fuel assembly rod power peaking factor of 3.04 is now referenced in the Technical Specifications as -

uel] as the 3.08 value for 7x7 fuel assemblies. The reduced fuel i pellet diameter of the 8x8 fuel assembly rods reduces the fuel time i constant and necessitates appropriate changes in the Technical Specif1-- t ca tio ns . We have previously reviesed(20) the performance characteristics

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j-l of the 8x8 fuel assemblics, including the effects of unpowered center rods, and abnormal. trensients (refer to item 1- of this evaluation) and '

l have concluded that reactor operation with 8x8 reload fuel asseablies in the Monticello core as well as 7x7 fuel assemblics is acceptable.

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Standby Gas Treatment Sys tem j The proposed technical specifications , we agree, clarify intentions and eliminate inconsis tencies bet

een plant charcoal filter _ tes t *

! capability and shop test capability uith reference to aerosol particle size. The changes involve changing _ the dioctyl,phtalate. particle

. size from 0.3 micron to 0.7 nicron and should be made. This change

, is acceptable based on Regulatory Guide 1.52.

Reactor Vessel Temperature Measurements

_ Temperatures on the outside surface of the reactor vessel are dependent i ~~ on the. coolant 'tacperature, cetal thickness, outside insulatica, and

, drywell ambient temperature. Since core coolant tenperature is the i only important_ variable; i.e., during heatup and-cooldown, the surface-

- ~ temperatures-Vill . change in a time dependent manner whenever loop temperatures change. - Having determined the relationship of the _ vessel -

surface temperatures to changes in loop temperature and verified that a the relationship is _ constant by repeated censurements ;during many pri-mary. coolant system heatup and cooldown operations, there is no need to' continue' this verification indefinitely. The data collected have been used to assure that_the rate of temperature change during normal heatup Eor cooldown of the reactor coolant system is not excessive.

Loop temperature restrictions , as specified in the Technical Specifi-Leations, are-adequate to guard against damage-due to rapid temperature i

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3 changes . We, therefore, concur that the requirement for reactor vessel shell and flange thermocouple measurements of Technical Specification

4.6. A.1 should be deleted. This change, we have concluded, does not

aff ect reactor saf ety or the health and safety of the public.

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! CONCLUSION We have reviewed thanges to technical specifications involving coolant pressure relief, control rods, standby gas treatment sys tem, and reactor J vessel surf ace temperatures, and have concluded that the changes do_ not present significant hazards considerations since the probability of accidents is not increased and safety margins to design limits are l not decreased and the severity of forseeable consequences are not

increased. He have also concluded that there is reasonable ascurance l that the health and safety of the public util not be endangered by the above changes , the changes to the fuel densification limits, and the

, use of 8x8 reload 2 f uel assemblies. Accordingly, the changes, as i presented by NSP and modified by us , in the replacement pages for the

Technical Specifications should be cade.

, James J. Shea-

! Operating Reactors Branch #2 '

Directorate of Licensing 4

Dennis L. Ziemann, Chief l

Operating Reactors Branch #2

Directorate of Licensing-Date: MAY ; [qa i

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REFERENCES

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USP subu i and errata dated January 23, 1974, and Marc ~. 19.

9 74, respectively " Permanent Plant Changes to Accommodate

'quilibrium Core Scram Reactivity Characteristics".

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linutes of HSP and AEC-L representatives meeting regarding replace-ment of Safety Valves and Prompt Relief Trip - by Directorate of Licensing dated March 14, 1974.

a. Directorate of Licensing letter dated March 14, 1974 "Approva) of Plant Modifications for Fuel Cycle 3".

4. NSF stLmittal dated March 8,1974 " Supplement to January 23, 1974 Report".

5. USP requas t Zor fuel loading and testing authorication dated March 21, 1974.

6. Minutes of NSF and AEC-L aceting o- ch 22, 1974 - by Directorate of Licc T .2ted April 2, 1974.

7 NSP sub:aittal dat ed ovenber 19,1973 "Second Reload Submittal".

8. Directorate of Licensing re;cer t for additional inf ormation related to PRI dated April 4,1974.

9. Safe., aluation by the Directorate of Licensing related to insercicn af 8x8 Fuel Asceablies into Monticello Core and Testing at Reactor Power Levels Less chan 1% dated bbrch 30, 1974.

10. Safety Evaluation by the Directorate of Licensing related to 8xS Fuel Assemolles - dated April 8, 1974.

11. NS? preposal to cha 2;c Technical Specifications to allow f or Cx8 F reload fuel dated February 27, 1974.

12. NcP proposal to changu Technical Specifications related to fuel densification - dated i'ebruary 28, 1974.

13. NSP preposal to change technical specifications related to sub-stitution of safety / relief valves for four spring-loaded safety velves and other PRI sys tem changes - dated March 1,1974.

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l 14. NSP response to AEC PRT concerns - dated April 26, 1974.

1 i 15. USP submittal dated September 13,1973 " Proposed Change in 7

Safety Valve Set Point" and Directorate of Licenslag Change

No. 10 dated October 2, 1973.

i 16. Directorate of Licensing Change No. 8 - Control Rod Scram Time

! Change - dated July 2, 1973.

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! 17. " Analytical Methods of Plant Transient Evaluations for the General .

l. Electric Boiling Uater Reactor" NEDO-10802 by GE dated February

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j 1973.

i i 18. NSF submittal dated February 8,1974 " Calculations Pertaining ,

! to the Densification Effects on 8x8 Fuel"'.

19. Directorate of Licensing Report - " Supplement 1 to the Tt.Jinical Report on Densification of General Electric Reactor Fuels" -

dated December 14, 1973.

! 20. Directorate of Licensing " Technical Report on the General Electric l Company 8x8 Fuel Assembly" - dated February 5,1974.

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