ML20195K090

Forwards Suppl to PWR Loop Ser:Use of Large Circulating Pump, Safety Evaluation of Unreviewed Safety Question

ML20195K090
Person / Time
Site:	MIT Nuclear Research Reactor
Issue date:	11/28/1988
From:	Bernard J MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20195K096	List: ML20195K090 ML20195K116
References
NUDOCS 8812050134
Download: ML20195K090 (9)

Text

I>. .

p%w

> 'l

.v.

• g c NUCLEAR REACTOR LABORATORY L is A.E 1

AN INTERDEPARTMENTAL CENTER OF MASSACHUSETTS INSTITUTE OF TECHNOLOGY O.K. HARUNG 138 Atany Street Cambridge, Mass. 02139 L. CLARK, JR.

Director (617)253- 4202 Director of Reactor Operations November 28, 1988 tT. S . Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D.C. 20555 Subj ect : Evaluation of an Unreviewed Safety Question, 10 CFR 50.59(b)(2) MIT Reactor License No. R-37, Docket No. 50-20 Centle.aen:

The Massachusetts Institute of Technology Nuclear Reactor Labora-tory is forwarding herewith a revision to a safety evaluation by the MIT Research Reactor Staff providing the bases for the determination that there are no unroviewed safety questions concerning the installa-tion and operation of an experiment on the MITR-II.

The experiment is a pressurized coolant chemistry loop which will be installed in the reactor core and which is described in the "Safety Evaluation Report f ar the PWR Coolant Chemistry Loop (PCCL)",

Report No. MITNRL-020, dated February 13, 1987 and its Supplement dated April 21, 1988. That report together with the safety evaluation performed by the MITR Staf f was subnitted to the U.S. Nuclear Regula-tory Commission on April 21, 1988. Enclosed here are a supplemetit to that report entitled, "Supplement to PWR Loop SER: Use of Large Circulating Pump" and the asr,oe f ated safety review. The purpose of the FCCI. enper f rie nt is to ir.vestigate the formation, transport, and deposition of radic4:tive crud in a carefully controlled loop that will simulate pressurized water reactor conditions, all wi th the objective of learning now to reduce radiation expcuures to personnel during maintenance work on these reactors.

Because experiments of this type are not described in the "Safety Analysis Report for the MIT Research Reactor (MITR-II)", Report No.

MITNE-Il5, October 22, 1970, as amended, a safety review of the exper-iment has been conducted by the project pe rs onnel , the reactor staff, and the MIT Reactor Safeguards Committee, including a safety evalua-tion as to the existence of any unreviewed safety questions (Safety Review #0-86-9, dated April 21, 1988 previously f orwa rded and SR

1. 0-88-5, dated September 9, 1988, copy enclosed). No unreviewed safety questions have been identified.

In accordance with 10 CFR 50.59(b)(2), evaluations of unreviewed safety questions have routinely been reported to NRC in the annual report required by paragraph 7,13.5 of the MITR-II Technical Specif f-cations, Facility License No. R-37. Because of the unusual nature of o As12050134 881123 FDR ADOCK 050000;O p PDL

o I USNRC Page 2 this experiment, we are reporting these safety evaluations required by 10 CFR 50.59 (b)(1) as they are performed rather than wait until the next MITR-II Annual Report (August 1989).

Our schedule now calls for the initial installation of this experiment in the reactor to occur in December 1988.

If you should have any questions regarding the evaluation or any of the information furnished, we request that you contact us as soon as possible.

Sincerely, rk R bP%

John A. Bernard , Ph.D Director of Reactor Operations Nuclear Reactor Laboratory JAB /crh

Enclosures:

Safety Review #0-8 8'- 5 , Sept. 9, 1988 which includes the Supplement to the Safety Evaluation Report, Oct. 24, 1988 cc: MITRSC USNRC - Region I Chief, Reactor Projects Section IB USNRC - Region I Project Inspector, Section ID USNRC - Resident Inspector, Pilgrim Nuclear Station i

SR #0-88 Supplement to PWR Loop SER ,

Use of Larne Circulatina Pump

Background

1. On April 5, 1988, the Special Subcomrnittee of the MIT Reactor Safeguards Committee, acting with the authority of the full MITRSC, approved the following documents:

(a) Safety Evaluation Report (SER) for the PWR Coolant Chemistry Loop (PCCL), MITNRL-020 dated February 13, 1987 and a supple-ment to that SER dated April 19, 1988.

(b) Safety Review #0-86-9 which concluded that there was no unreviewed safety question associated with the installation of the PCCL.

(c) Letter to the U.S. Regulatory Commission f orwarding SRf 0-86-9 and the PCCL SER.

(

Subsequent to that meeting, the above mentioned material was sub-mitted to the U.S. Nuclear Regulatory Commistion in accordance with the provisions of 10 CTR 50.59.

2. One of the provisions of the PCCL approval concerned hydrostatic testing of the PCCL thimble and Jacket. It was stated, "The thimble and jacket are designed for, and will be hydrostatically tested at, 750 psi. They are protected by redundant pressure relief valves set at 30-100 psi and a rupture disk rated for about 500 p s f. " . During thc period 20-23 June 1988, the it . S .

Nuclear Regulatory Corut.1 m a ton conducted a routine inspection which identified no vinistions. However, a concern was raised relative to the hydrostatic testing of the PCCL. Given in Table One is the relevant excerpt from the NRC report together with the technical specification of concern. The NRC position was that the hydrer.tatic tests should be conducted to t:41ce the maximum expacted pressure or 910 psi.

3. In late June 1988, the FCCL experimenters proposed two changes to the PCCL. These were (1) a reanalysis of the calculation of the pressure transient resulting frem a loss of PCCL coolant incident and (2) a revision of the PCCL design to allow utiliza-tion of an out-of-pile pump.

Description of Change

1. The proposed changes are described in the "Supplement to the PWR Loop SER: Use of Large Circulating Pump" dated 15 September 1988, a copy of which is attached as Appendix A to this safety review.

SR#-0-88-5 SEP 09 1988

Safety Evaluation

1. ta rste Pump option: Hazards associated with the large pump options are identified ass (a) N-16 Radiation on the Reactor Top - This will be addressed by oncasing the pump in a lead-lined shield and by requiring surveys by HITR-RPO as part of the initial startup sequence for the FCCL.

(b) Ex-Core Laak - Leak tapes or probes will be used to detect minor leaks. Large leaks vou.'d be apparent as loss of PCCL coolant inventory. The use of local ventilation plus insula-tion wrapping the pump and its piping will preclude both an airborne radiation hazard and the possibility of spraying personnel with hot PCCL coolant.

(c) In-Core Leak - Large leaks would be detected as loss of PCCL inventory. Small leaks would be detectable only over time.

This means that, in the event of a small leak the MITR's pri-mary coolant would be contaminated by the PCCL's coolant.

This would be undesirable because the former is maintained slightly acidic (for aluminum) and boron-free. The latter is maintained basic (for sreel) and contains boron. Protection is provided by the MITR's coolant conductivity monitoring system and by the fact that the intermixing of PCCL and MITR coolant will have little effect because of the differences in volumes. PH 5.2.12, "High Specific Conductance - Primary System" will be revised prior to PCCL installation to require i checking of PCCL operation in the event of an increate in conductivity. ( F2.t,.*, The boren in the FCCL is depleted in boron-10, the toutron absorbing isotope. Hence, leakage would havs little, if any, reactivity effect.)

(d) Loss of Shield Co,q Qnt - The ex"core pump will require use of the shield coolar.t system as do many other experiments. 1 Existing alarms on the shield system will provide adequate warning f.o the operator of any loss of shield soolant inven- ,

i tory.

Given the above, it is ccncluded that no unreviewed safety ques-tion exists relative to the use of an ex-core pump. The hazards identifiud (N-16, leakage, coolant contamination, and loss of shleid coolant inventory) are not new. These have all been i addressed in similar form as part of previous experiments.

Similarly, the probability of these hazards is not increased.

Each is being addressed in a manner consirtent with past conser-vative practices, yinally, no margin of safety as defined in the basis of any technical specification has been reduced.

2. Hydrostatle Tests As described in Section 4 of the SER supplement, the PCCL loss of SRd-0-88-5 SEP 09 1988 i

coolant accident has been reanalyzed. Also, the necersary hydro-static tests have now been performed. The following inf o rmation is relevant:

(a) Relative to the concerns raised by the U.S. Nuclear Regula-tory Commission about the requirements of technical specifi-cation #6.1.3a, a prototype of the in-core section of the PCCL has been tested to 1000 psi. This more than satisfies the test of 970 psi recommended during the NRC inspection as being necessary. (Note: The technical specification speci-fica 11y authorizes ' prototype' testing.)

(b) The thimble burst disk and its associated pressure relief valve have baen moved out of core allowing the use of larger diameter pipes and hence significant reductions in pressure.

Also, the loss of PCCL coolant accident analysis has been redone to improve its accuracy. The following figures now apply:

Pressure Relief /alves 20 psi Burst Disk: 65 psi Thimble Hydrostatic Test Pressura: 150 psi Maximum Pressure if No Relief: 150 psi Prototype Test Pressure: 1000 psi Every thimble will be tested to 150 psi, more than twice the maximum possible pressure of 65 psi. (Note: Burst disk rupture at 65 psi should limit pressure to this value.)

Given the above it is concluded that no unreviewed safety question exists relative to the revised analysis of the PCCL loss of coolant accident. There is no increase in probabil-ity of an analyzed accident , no new type of accident, and no decrease in the margin of safity as defined in the basin of a I technical specification. In fact, safety margins have been increased. Also, concerns raised by the U.S. Nuclear Regula~

tory Commission during fespection #88-02 have besn fully addressed.

3. As originally approved, 10 scf of hydrogen could be stored within the containment for purpoJes of this experiment. The experi-menter has now requested a limit of 20 scf because the smallest commercial cylinders are 15 set. It is felt that use of a stan-dard commercial product is far safer than the alternative which would be to transf er hydrogen f rom the commercial container to some other holder. It is further noted that the in-core inven-tory of hydrogen will not be increased. Also, the total energy associated with 20 scf of hydrogen is small. Given these facts it is concluded that no unreviewed safety question exists rela-tive to the hydrogen storage issue. No new type of accident 16 created, there is no increase in probability of an analyzed acci-SRd-0-88-5 SEr 09 1986

s

-4 i

dent, and no margin of safety as defined in the basis of a technical specification is reduced. If anything, this change 1 will lessen the probability of a problem through uso of a  ;

standard, cosasercial product.

l l

1 l

i t

l l

I t

i i

t l

3y.o.33 5 m 09 1988

t Table One I A. Excerpt from U.S. Nuclear Reaulatory Commission Report #50-20/

88-02 i The last experiment of this category involves the simulation of [

commercial nuclear facility operating loops to develop strategies t to minimize activities. This will involve operation of a scale [

model loop within the reactor core region while operating, and t its subsequent removal and dissection. Although the physical size of the loop will be scaled down, pressures, temperatures and  !

flow velocity will be equal to that of the loop being simulated.

The licensee is close to approving the installation of the first ,

experiment of this type which will simulate a Westinghouse Pres-  ;

surized Water Reactor Chemistry Loop (PCCL). The licensee has determined that this experiment does not involve an URSQ or i require a Technical Specification change. The licenses further I intends to submit a description of the experiment to the NRC per [

10CyR50.59 prior to its installation rather than wait to include  !

it in the annual report to the NRC. The inspector reviewed the safety analysis for this experiment and found it to be designed i within the boundaries of the Technical Specifications with the following exception:

MIT T'chnical Specification 6.1.3 requires that experiment l materials placed in the core region which could react to cause a

[

pressure spike be encapeulated within a capsule prototype tested '

to 2 times the maximum expected pressure spike. The maximum [

pressure resulting from a rupture of the FCCL is calculated to be

(

485 psi. The aluminum thimble which will be used to encapsulate i the FCCL was described in the Safety Analysis as having a "Proof I Pressure" of 750 psi which is significantly lower than the h required 970 psi.

Thia concern was id e rat if i ed to f acility management. Although this specification is included in a riection of the Technical Specifiaations involving chemical effects of experiments, the licen se agreed with the inspector's pcsition that thin specifi.

cat' i should also be applied to direct mechanical effects. The ite.nsee agread to re addreas ?.his issue prior to final approval of the exp1riment. This item is unresolved pending final 11cen-see disposition (80-02-01). No other inadequacies were identi-fled.

B. Technical Speelfication 46.1.3a

3. Chemical Effects

a. Metastable or other materials that could react to create SR#-0-88-5 SEP 09 1988

l Table One (cont.)

a rapid pressure rise shall be encapsulated. The capsule shall be prototype tested under experimental conditions to demonstrate that it can contain without failure an energy release equivalent to at least twice the material to be irradiated or at least twice the pressure that l could be expected from any reaction of these materials.

These tests must also include effects of any f ragmen.i e which may be generated. If a change in experimental con-ditions could result in a greater potential for failure than design experimental conditions, the capsule shall also be tested under these changed conditions. In addi-tion, the quantity of material should be limited such that if the maximum calculated energy release shoulo occur, significant damage to the reactor core will not result, assuming the material is not encapsulated.

l l

l f

l i

SRf-0-88-5 SEP 09 1988

- . . . _ _ _ - - - - ~ ~ - . . ..- .. _ - - - . .

- li

.y.

Anoendix A Supplement to the FWR Loon SIhr Use of Larne Circulatina Puno O